NGNP and Hydrogen Production Conceptual Design Study

NGNP-NHS 100-RXBLDG September 2008Revision 0

NGNP and Hydrogen Production ConceptualDesign Study

Reactor Building Functional and Technical Requirementsand Evaluation of Reactor Embedment

APPROVALS

Function Printed Name and Signature DateAuthors Peter Wells

Shaw Environmental & Infrastructure

Karl FlemingTechnology Insights

September 16,2008

Reviewerpp Michael CorreiaPebble Bed Modular Reactor (Pty) Ltd.

September 16,2008

ReviewerFred SiladyTechnology Insights

September 15,2008

ReviewerDan MearsTechnology Insights

September 15, 2008

ApprovalEdward BrabazonShaw Environmental & Infrastructure

September 16,2008

Westinghouse Electric Company LLCNuclear Power PlantsPost Office Box 355

Pittsburgh, PA 15230-0355

©2008 Westinghouse Electric Company LLCAll Rights Reserved

NGNP-NHS 100-RXBLDG Rev 0 NGNP Conceptual Design StudyReactor Building Functional and Technical

Requirements and Evaluation of Reactor Embedment

NGNP-NHS 100-RXBLDG 0 Final 091608.doc September 15, 20082 of 186

LIST OF CONTRIBUTORS

Name and Organization Date

Michael DavidsonShaw Environmental & Infrastructure

September 2008

William MartinShaw Environmental & Infrastructure

September 2008

Steve VigeantShaw Environmental & Infrastructure

September 2008

Robert WilmerShaw Energy & Chemicals

September 2008

Fred SiladyTechnology Insights

September 2008

David DillingTechnology Insights

September 2008

Alfred TorriTechnology Insights

September 2008

Lori MascaroTechnology Insights

September 2008

Dirk UysPebble Bed Modular Reactor (Pty) Ltd.

September 2008

Roger YoungPebble Bed Modular Reactor (Pty) Ltd.

September 2008

Johan StraussPebble Bed Modular Reactor (Pty) Ltd.

September 2008

Wilma van EckPebble Bed Modular Reactor (Pty) Ltd.

September 2008

James WintersWestinghouse Electric Company

September 2008




BACKGROUND INTELLECTUAL PROPERTY

Section Title Description

A3 Evaluation of Reactor Building Pressure Response andRadiological Retention Capability Reference [14]

Radionuclide retention scoping calculations and source term assumptions

Reactor Building pressure and temperature calculations

B1.2 DLOFC transient Analysis of RCCS Standpipes, PBMR Calculation TA00024/A issued 5 June 2004. Reference [27]

Scoping analysis of temperature transient without RCCS operating.

B1.2 Heat transfer calculation for Reactor Embedment Task Design Information Transmittal DIT001254 taking information from Calculation T001284 August 2008. Reference [28]

Scoping analysis of differences of heat transfer from reactor to exterior with soil and air.




REVISION HISTORY

RECORD OF CHANGES

Revision No. Revision Made by Description Date

0 Initial issue September 2008

DOCUMENT TRACEABILITY

Created to support the following Document(s)

Document Number Revision

None None




TABLE OF CONTENTS

LIST OF TABLES...................................................................................................................................... 9

LIST OF FIGURES.................................................................................................................................. 10

ACRONYMS ............................................................................................................................................ 12

SUMMARY AND CONCLUSIONS ......................................................................................... 16INTRODUCTION....................................................................................................................... 27OBJECTIVE AND SCOPE ....................................................................................................... 28ORGANIZATION OF REPORT .............................................................................................. 29PART A: REACTOR BUILDING TECHNICAL AND FUNCTIONAL REQUIREMENTS

AND EVALUATION OF ALTERNATIVE CONFINEMENT STRATEGIES.......... 30A1 DEFINITION OF TECHNICAL AND FUNCTIONAL REQUIREMENTS.............. 30A1.1 TECHNICAL APPROACH TO DEFINITION OF REQUIREMENTS............................... 30

A1.2 NGNP TOP LEVEL REQUIREMENTS .................................................................................. 32

A1.3 REVIEW OF APPLICABLE NRC REQUIREMENTS.......................................................... 33

A1.4 DEFINITION OF SAFETY AND LICENSING REQUIREMENTS..................................... 36

A1.4.1 ROLE OF REACTOR BUILDING IN NGNP SAFETY DESIGN APPROACH.... 36

A1.4.2 IMPLICATIONS OF RI-PB LICENSING APPROACH .......................................... 41

A1.4.3 PRELIMINARY LICENSING BASIS EVENTS ........................................................ 42

A1.5 DEFINITION OF PHYSICAL SECURITY REQUIREMENTS ........................................... 44

A1.6 REACTOR BUILDING REQUIREMENTS SUMMARY...................................................... 46

A2 DEFINITION OF ALTERNATIVE REACTOR BUILDING DESIGN CONCEPTS............................................................................................................................................. 54

A2.1 ALTERNATIVE 1a UNFILTERED AND VENTED .............................................................. 60

A2.2 ALTERNATIVE 1b VENTED AND UNFILTERED WITH RUPTURE PANELS............. 62

A2.3 ALTERNATIVE 2 PARTIALLY FILTERED AND VENTED WITH RUPTURE PANELS....................................................................................................................................................... 63

A2.4 ALTERNATIVE 3a FILTERED AND VENTED WITH RUPTURE PANELS................... 65

A2.5 ALTERNATIVE 3b FILTERED AND VENTED WITH RUPTURE PANELS AND EXPANDED VOLUME.............................................................................................................. 66

A2.6 ALTERNATIVE 4a PRESSURE RETAINING WITH INTERNAL RUPTURE PANELS

....................................................................................................................................................... 67

A2.7 ALTERNATIVE 4b PRESSURE RETAINING WITH INTERNAL RUPTURE PANELS AND EXPANDED VOLUME .................................................................................................... 68




A3 EVALUATION OF REACTOR BUILDING PRESSURE RESPONSE AND RADIOLOGICAL RETENTION CAPABILITY.......................................................... 70

A3.1 CHARACTERIZATION OF FISSION PRODUCT TRANSPORT ...................................... 70

A3.2 TECHNICAL APPROACH ....................................................................................................... 72

A3.3 KEY ASSUMPTIONS ................................................................................................................ 77

A3.4 BOUNDARY CONDITIONS ..................................................................................................... 78

A3.5 CHARACTERIZATION OF INITIAL RADIONUCLIDE INVENTORIES ....................... 78

A3.6 MATRIX OF RADIOLOGICAL EVALUATION CASES..................................................... 79

A3.7 CHARACTERIZATION OF RELEASES FROM FUEL DURING DLOFC....................... 81

A3.8 EVALUATION OF SITE BOUNDARY DOSES FOR THE NO REACTOR BUILDINGCASE ............................................................................................................................................ 83

A3.9 EVALUATION OF SITE BOUNDARY DOSES FOR REACTOR BUILDING ALTERNATIVES ....................................................................................................................... 84

A3.10 SUMMARY OF RESULTS FOR REACTOR BUILDING PRESSURE RESPONSE......... 92

A3.11 AIR INGRESS ............................................................................................................................. 94

A3.12 CONCLUSIONS AND RECOMMENDATIONS FOR FURTHER STUDY........................ 97

A4 INTEGRATED EVALUATION OF REACTOR BUILDING DESIGN CONCEPT ALTERNATIVES.............................................................................................................. 99

A4.1 TECHNICAL APPROACH TO EVALUATION .................................................................... 99

A4.2 EVALUATION OF REACTOR BUILDING ALTERNATIVES......................................... 100

A4.2.1 EVALUATION OF NORMAL OPERATION CRITERION.................................. 101

A4.2.2 EVALUATION OF INVESTMENT PROTECTION CRITERION....................... 101

A4.2.3 EVALUATION OF SAFETY CRITERION.............................................................. 102A4.2.3.1 Evaluation of Helium Pressure Boundary Break Response......................102A4.2.3.2 Evaluation of Radiological Retention Criterion.........................................102A4.2.3.3 Evaluation of Seismic Capability.................................................................103

A4.2.4 EVALUATION OF PROCESS HAZARDS CRITERION ...................................... 104

A4.2.5 EVALUATION OF PHYSICAL SECURITY AND AIRCRAFT CRASHCRITERIA .............................................................................................................. 104

A4.2.6 EVALUATION OF CAPITAL AND OPERATING COSTS................................... 105

A4.2.7 EVALUATION OF LICENSABILITY CRITERION.............................................. 105

A4.2.8 SUMMARY OF EVALUATION RESULTS............................................................. 106

A4.2.9 EVALUATION SUMMARY OF ALTERNATIVE REACTOR BUILDING CONCEPTS .............................................................................................................. 108

A5 OPEN ISSUES AND ADDITIONAL ENGINEERING STUDIES............................. 110PART B: EVALUATION OF REACTOR EMBEDMENT ................................................ 112




B1 LIST OF ISSUES RELEVANT TO REACTOR EMBEDMENT .............................. 114B1.1 OPERATIONAL NEEDS......................................................................................................... 114

B1.2 HEAT DISSIPATION TO THE ENVIRONMENT............................................................... 126

B1.3 WATER TABLE EFFECTS .................................................................................................... 128

B1.3.1 FLOTATION .............................................................................................................. 129

B1.3.2 WATER PRESSURE ON EMBEDDED WALLS..................................................... 129

B1.3.3 PERMANENT DEWATERING SYSTEMS ............................................................. 129

B1.3.4 WATERPROOFING ................................................................................................... 130

B1.3.5 FLUID-APPLIED WATERPROOFING MEMBRANE.......................................... 130

B1.3.6 CONSTRUCTION DEWATERING .......................................................................... 131

B1.4 GEOTECHNICAL CONSTRAINTS AND FOUNDATION PERFORMANCE................ 132

B1.4.1 TEMPORARY EXCAVATION SUPPORT.............................................................. 132B1.4.1.1 Excavations in Soil ........................................................................................132B1.4.1.2 Excavations in Rock......................................................................................133

B1.4.2 LATERAL EARTH PRESSURES ............................................................................. 134B1.4.2.1 Static Lateral Earth Pressure ......................................................................134B1.4.2.2 Dynamic Lateral Earth Pressure.................................................................134

B1.4.3 BACKFILL AROUND EMBEDDED STRUCTURE............................................... 135

B1.4.4 ALLOWABLE BEARING PRESSURE .................................................................... 135

B1.4.5 MODULUS OF SUBGRADE REACTION ............................................................... 135

B1.4.6 SLIDING STABILITY ................................................................................................ 136

B1.4.7 STRUCTURE OVERTURNING................................................................................ 136

B1.4.8 STRUCTURE SETTLEMENT................................................................................... 137

B1.4.9 INPUT TO SOIL-STRUCTURE INTERACTION ANALYSIS.............................. 137

B1.5 CONSTRUCTION CONSIDERATIONS ............................................................................... 138

B1.6 COST CONSIDERATION....................................................................................................... 139

B1.6.1 CAPITAL COST .......................................................................................................... 139

B1.6.2 LIFE CYCLE COST.................................................................................................... 142

B1.7 MALEVOLENT HAZARDS.................................................................................................... 143

B1.8 NATURAL PHENOMENON HAZARDS .............................................................................. 144

B 1.8.1 INTRODUCTION........................................................................................................ 144

B1.8.2 TORNADO HAZARDS ............................................................................................... 145

B1.8.2 FLOODING HAZARDS.............................................................................................. 146

B1.9 NATURAL GEOLOGICAL HAZARDS ................................................................................ 146




B1.9.1 TECTONIC AND SEISMIC HAZARDS................................................................... 146B1.9.1.1 Seismic Classification....................................................................................146B1.9.1.2 OBE Design Basis..........................................................................................147B1.9.1.3 Ground Motion Characteristics...................................................................147B1.9.1.4 Analysis & Design .........................................................................................148B1.9.1.5 Soil Structure Interaction.............................................................................149

B1.9.2 NON-TECTONIC SURFACE DEFORMATION..................................................... 150

B1.9.3 VOLCANIC HAZARDS.............................................................................................. 150

B1.9.4 EFFECTS ON EMBEDMENT STUDY..................................................................... 151

B1.10 CHEMICAL RELEASE, EXPLOSIONS, MANMADE HAZARDS ................................... 151

B1.10.1TOXIC CHEMICAL RELEASES.............................................................................. 151

B1.10.2ASPHYXIANT GAS RELEASES .............................................................................. 153

B1.10.3FLAMMABLE RELEASES........................................................................................ 154

B1.10.4OXYGEN RELEASES................................................................................................. 155

B1.10.5SUMMARY .............................................................................................................. 156

B2 EVALUATION AND RECOMMENDATION REGARDING REACTOREMBEDMENT................................................................................................................. 157

B.2.1 OBJECTIVE RANKING.......................................................................................................... 157

B2.2 LIST OF ALTERNATIVE SITES CONSIDERED ............................................................... 159

B2.3 ALTERNATIVE REACTOR EMBEDMENT CONCEPTS................................................. 159

B2.4 SCORING CRITERIA ............................................................................................................. 159

B2.5 ALTERNATIVE RANKING SUMMARY ............................................................................ 166

B2.6 ROLE OF EMBEDMENT TO SATISFY T&FRS..................................................................... 167

B2.7 RECOMMENDATIONS .......................................................................................................... 175

B3 OPEN ISSUES AND ADDITIONAL R&D AND ENGINEERING STUDIES ......... 176REFERENCES.......................................................................................................................... 177BIBLIOGRAPHY................................................................................................................................... 181

DEFINITIONS........................................................................................................................................ 182

REQUIREMENTS ................................................................................................................................. 183

ASSUMPTIONS ..................................................................................................................................... 184

APPENDICES ........................................................................................................................................ 186

APPENDIX A: 90 PERCENT DESIGN REVIEW PRESENTATION TO BEA ... 186




LIST OF TABLES

Table 1 Alternative Reactor Building Strategies for Performing Safety Functions ................ 19Table 2 Evaluation of Embedment Alternatives for Rock and Soil Sites................................ 24Table 3 Elements of NGNP Risk-Informed Performance-Based Licensing Approach........... 33Table 4 Key Differences between NGNP and NUREG-1860 RI-PB Approaches.................. 35Table 5 Preliminary LBEs Involving HPB Leaks and Breaks................................................. 43Table 6 Preliminary Technical and Functional Requirements for NGNP Reactor Building... 49Table 7 Alternative Reactor Building Strategies for Performing Safety Functions ................ 57Table 8 NGNP Initial Conditions for Model ........................................................................... 76Table 9 Top Level Regulatory Criteria used to Evaluate Site Boundary Doses...................... 78Table 10 Initial Release Source Terms for I-131 and Cs-137 Radio-nuclides .......................... 79Table 11 Inputs Used for the RB Model to Simulate the Different RB Alternatives ................ 81Table 12 Dose Results for the RB Alternatives with a 10 % RB Vent Volume........................ 85Table 13 Dose Results for the RB Alternatives with a large RB Vent Volumes....................... 86Table 14 Detailed Results for RB Alternative 1a ...................................................................... 87Table 15 Design Basis Peak Pressures for RB Design Alternatives.......................................... 93Table 16 Peak Pressures for the 1000 mm SHTS Hot Leg Break ............................................. 94Table 17 Evaluation Criteria and Relative Weights................................................................... 99Table 18 Evaluation Scores for Individual Criteria ................................................................. 107Table 19 Total Evaluation Scores for Reactor Building Alternatives ..................................... 107Table 20 Reactor Building Systems......................................................................................... 121Table 21 Cost Analysis ............................................................................................................ 141Table 22 Hypothetical Reactor Building Seismic Accelerations............................................. 150Table 23 Objective Ranking and Relative Weights ................................................................. 157Table 24 Alternative Ranking Rock Sites (Deep Water Table) Minimum Embedment ......... 160Table 25 Alternative Ranking Rock Sites (Deep Water Table) Partial Embedment ............... 161Table 26 Alternative Ranking Rock Sites (Deep Water Table) Full Embedment ................... 162Table 27 Alternative Ranking Soil Sites (High Water Table) Minimal Embedment .............. 163Table 28 Alternative Ranking Soil Sites (High Water Table) Partial Embedment ................. 164Table 29 Alternative Ranking Soil Sites (High Water Table) Full Embedment ..................... 165Table 30 Design Alternatives Ranking .................................................................................... 166Table 31 Role of Embedment in Satisfying Reactor Building Functions and Requirements.. 168




LIST OF FIGURES

Figure 1 Site Boundary Dose (TEDE) vs. HPB Break Size for LBEs Involving DLOFC for Alternative Reactor Building Concepts with Comparable Volumes............................ 20

Figure 2 Flow Chart of Part A: Development of Reactor Building Requirements and Evaluation of Alternative Concepts ............................................................................. 31

Figure 3 Key Elements of Plant Capabilities Supporting Defense-in-Depth ............................ 37Figure 4 Top Down Development of NGNP Safety Functions .................................................. 38Figure 5 NGNP Reactor Building in Preconceptual Design Report.......................................... 47Figure 6 Reactor Building Alternative 1a – Unfiltered and Vented ........................................... 61Figure 7 Reactor Building Alternative 1b – Unfiltered and Vented with Rupture Panels ......... 62Figure 8 Reactor Building Alternative 2 – Partially Filtered And Vented With Rupture

Panels............................................................................................................................ 63Figure 9 Reactor Building Alternative 3a - Filtered and Vented With Rupture Panels ............. 65Figure 10 Reactor Building Alternative 3b Filtered and Vented With Rupture Panels and

Expanded Volume ........................................................................................................ 66Figure 11 Reactor Building Alternative 4a Pressure Retaining With Internal Rupture Panels ... 67Figure 12 Reactor Building Alternative 4b Pressure Retaining With Internal Rupture

Panels and Expanded Volume...................................................................................... 69Figure 13 Helium Flow Path Through NGNP Reactor Pressure Vessel ..................................... 71Figure 14 NGNP Average Fuel Temperature during DLOFC..................................................... 76Figure 15 RB Alternative 2 Used as Basis for the RB Analytical Model.................................... 80Figure 16 Core Fuel Time at Temperature Distribution during DLOFC..................................... 82Figure 17 I-131 and Cs-137 Cumulative Activity Release from Fuel during DLOFC .............. 83Figure 18 Dose Behavior vs. Break Size for RB Alternative 1a with 10% RB Vent Volume ... 85Figure 19 Time dependent release of I-131 for a 3 mm Break in Alternative 1a with a

10% RB Vent Volume.................................................................................................. 88Figure 20 Time dependent release of I-131 for a 2 mm Break in Alternative 1a with a

10% RB Vent Volume.................................................................................................. 89Figure 21 Comparison of the I-131 Release for the 2 mm Break and the 3 mm Break for

Alternative 1a ............................................................................................................... 90Figure 22 Dose vs. Break Size for the RB Alternatives with 10% RB Vent Volume................. 91Figure 23 Thyroid Dose for the alternative Designs with a 10% RB Vent Volume for the

3 mm Break Size .......................................................................................................... 92Figure 24 Flow Chart of Part B: Evaluation of Reactor Embedment ........................................ 113Figure 25 Access to Reactor Building ....................................................................................... 116Figure 26 Access to Reactor Building at Main Equipment Elevator......................................... 117Figure 27 Access to Reactor Building at Lowest Level in the Building ................................... 117Figure 28 Elevation Requirements of Major SSC With Respect to the Reactor ....................... 118Figure 29 NGNP Reactor Building Layout ............................................................................... 120Figure 30 Top View of Reactor Cavity Schematically Showing the Added Layers ................. 127Figure 31 Maximum Concrete Temperature at a Height Corresponding to the Middle of

the Pebble Bed............................................................................................................ 127




Figure 32 Average Concrete Temperature at a Height Corresponding to the Middle of the Pebble Bed............................................................................................................ 128

Figure 33 Estimated Capital Cost of Key Elements .................................................................. 142Figure 34 Reactor Building Seismic Accelerations ................................................................... 149




ACRONYMS

Abbreviation or Acronym

Definition

ACI American Concrete Institute

ACRS Advisory Committee on Reactor Safeguards

AEC Atomic Energy Commission

ALARA As Low As Reasonably Achievable

ALOHA Areal Locations of Hazardous Atmospheres

ANPR Advance Notice of Proposed Rulemaking

ANSI American National Standards Institute

AOO Anticipated Operational Occurrence

ASCE American Society of Civil Engineers

ASME American Society of Mechanical Engineers

ASTM American Society of Testing and Materials

AVR Arbeitsgemeinschaft Versuchsreaktor

BDBE Beyond Design Basis Event

BLEVE Boiling Liquid Expanding Vapor Explosion

BOD Basis of Design

CBCS Core Barrel Conditioning System

CCS Core Conditioning System

CEPPO Chemical Emergency Preparedness and Prevention Office

CFR Code of Federal Regulations

CIP Core Inlet Pipe

COL Combined License

COP Core Outlet Pipe

CP Construction Permit

DBA Design Basis Accident

DBE Design Basis Event

DBT Design Basis Threat

DC Design Certification





Definition

DCD Design Control Document

DEGADIS Dense Gas Dispersion

DEGB Double Ended Guillotine Break

DID Defense-In-Depth

DLOFC Depressurized Loss of Forced Cooling

DOE Department of Energy

DPP Demonstration Power Plant

DSA Deterministic Safety Analysis

EAB Exclusion Area Boundary

EPA Environmental Protection Agency

EPRI Electric Power Research Institute

EPZ Emergency Planning Zone

FHSS Fuel Handling and Storage System

GDC General Design Criteria

HPB Helium Pressure Boundary

HPF Hydrogen Production Facility

HPS Hydrogen Production System

HSS Helium Service System

HTGR High Temperature Gas Cooled Reactor

HVAC Heating, Ventilation, and Air Conditioning

IBC International Building Code

ICM Interim Compensatory Measures

IDLH Immediate Danger to Life and Health

IHX Intermediate Heat Exchanger

INL Idaho National Laboratories

LBE Licensing Basis Event

LWR Light Water Reactor

NGNP Next Generation Nuclear Plant

NHSB Nuclear Heat Supply Building

NHSS Nuclear Heat Supply System





Definition

NIOSH National Institute of Occupational Safety and Health

NOAA National Oceanographic and Atmospheric Administration

NPH Natural Phenomenon Hazard

NRC Nuclear Regulatory Commission

NSR Non Safety Related

OBE Operating Basis Earthquake

PAG Protective Action Guideline

PBMR Pebble Bed Modular Reactor

PCDR Pre-Conceptual Design Report

PCHX Process Coupling Heat Exchanger

PCS Power Conversion System

PHTS Primary Heat Transport System

PIRT Phenomenon Identification and Ranking Tables

PLOFC Pressurized Loss of Forced Cooling

PRA Probabilistic Risk Assessment

PRS Pressure Relief System

PSP Physical Security Plan

RB Reactor Building

RBVV Reactor Building Vent Volume

RCCS Reactor Cavity Cooling System

RDC Regulatory Design Criteria

RG Regulatory Guide

RIM Reliability and Integrity Management Program

RI-PB Risk Informed – Performance Based

SC Seismic Category

SC-H Safety Class - High

SC-L Safety Class - Low

SC-M Safety Class - Medium

SG Steam Generator

SGI Safeguards Information





Definition

SHTS Secondary Heat Transport System

SR Safety Related

SRM Staff Requirements Memorandum

SSC Systems, Structures and Components

SSE Safe Shutdown Earthquake

ST Special Treatment

T&FR Technical and Functional Requirements

TEDE Total Effective Dose Equivalent

TLRC Top Level Regulatory Criteria

TNT Trinitrotoluene

URD Utility Requirements Document

UVCE Unconfined Vapor Cloud Explosion




SUMMARY AND CONCLUSIONS

This study develops technical and functional requirements (T&FRs) for the NextGeneration Nuclear Plant (NGNP) and High Temperature Gas-cooled Reactor (HTGR) ReactorBuilding (RB), including consideration of reactor embedment. The following Summary and conclusions are organized into three parts. The first part (Part A) summarizes the overall technical and functional requirements for the reactor building and the evaluation of various design strategies for meeting these requirements. The second part (Part B) of the summary and conclusions identifies criteria and requirements relevant to determining the degree of reactor building embedment and evaluates and ranks embedment depth alternatives. The third part of the summary and conclusions focuses on the role that embedment can contribute to meetingT&FRs identified in the previous discussion.

Reactor Building Functions and Requirements (Part A)

A comprehensive set of technical and functional requirements for the NGNP Reactor Building was developed in this study for use in the Conceptual Design Phase to follow. These requirements were developed in a top down fashion starting with the NGNP Top Level Requirements identified in the Preconceptual Design Report (Reference [1]) and the following considerations:

• Role of the Reactor Building (RB) in the Pebble Bed Modular Reactor (PBMR) NGNP safety design approach

• Review of NRC requirements and guidance applicable to the Reactor Building including those for safety and physical security

• The Risk-Informed and Performance Based Licensing approach envisioned for the PBMR NGNP (References [3] through [6])

• Selection of Preliminary Licensing Basis Events (LBEs) for the NGNP

The reactor building is designed to house and protect the Systems, Structures, and Components (SSCs) within the Nuclear Heat Supply System (NHSS), which include, among other SSCs, the Reactor, Primary Heat Transport System (PHTS), portions of the Secondary Heat Transport System (SHTS), Helium Services System (HSS) and the Fuel Handling and Storage Systems (FHSS). In devising a comprehensive set of Reactor Building functions, the following categories of functions were considered:

• Safety Functions

o Required Safety Functions: those functions that are necessary and sufficient to meet the dose limits for Design Basis Events (DBE) and deterministically selected Design Basis Accidents (DBA)




o Supportive Safety Functions: all other functions that contribute to the prevention or mitigation of accidents and support the plant capabilities for defense-in-depth

• Physical Security Functions: functions to protect the reactor and vital equipment from design basis threats associated with acts of sabotage and terrorism

• Non-safety/security Functions: those functions necessary for plant construction, operation, maintenance, access, inspection, worker protection, and control of routine releases of radioactive material during normal operation

The RB shall perform the following required safety and security functions:

• House and provide civil structural support for the reactor and all the SSCs within the scope of the NHSS (building geometry/space)

• Resist all structural loads as required for required safety functions allocated to NHSSSSCs include those associated with:

o Prevention and mitigation of releases of radio-nuclides

o Control core heat generation

o Control of core heat removal

o Control of chemical attack by protecting the PHTS Helium Pressure Boundary (HPB) integrity

o Maintenance of core and reactor vessel geometry

o Maintenance of reactor building structural integrity

• Protect the SSCs within the NHSS that perform required safety functions from all internal and external hazards as identified in the LBEs

• Provide physical security of vital areas of the NHSS against acts of sabotage and terrorism

The RB shall perform supportive safety functions involving the retention of radio-nuclides released from the PHTS HPB and the limiting of air ingress following HPB leaks and breaks. These supportive safety functions are shared by the RB and the RB Pressure Relief System (PRS) and the Reactor Building HVAC System.

A preliminary set of Licensing Basis Events (LBEs) was selected to formulate the Reactor Building Technical and functional Requirements. A final set of LBEs shall be defined with input from a full scope PRA during conceptual and preliminary design stage. For this study LBEs are defined by applying engineering judgment based on reviews of the plant design per PCDR and experience with HTGR and specifically PBMR PRAs.




Categories of LBEs include:

• Internal Eventso HPB leaks and breaks in PHTS and SHTS

o Transients with and without reactivity addition

• Leaks and breaks in PHTS Heat Exchangers such as those in the Core Conditioning System (CCS) and Intermediate Heat Exchanger (IHX)

o Internal Plant Hazard Events

o Internal fires and floods

o Rotating machinery missiles

• External Plant Hazard Events o Hydrogen Process events

o Seismic Events

o Aircraft crashes and transportation accidents

o High winds and wind generated missiles

A preliminary set of LBEs involving leaks and breaks in the PHTS HPB was developed based on a review of similarities and differences of the NGNP design against other PBMR designs for which leak and break initiating event frequencies have been quantified. These LBEs were used to evaluate the Reactor Building capabilities for radiological retention and mitigation of temperature and pressure loads in the formulation of requirements.

Evaluation of Alternative Reactor Building Design Concepts (Part A)

A key result of this study was the evaluation of alternative Reactor Building concepts for the satisfaction of the building safety functions involving pressure relief for HPB leaks and breaks, retention of radioactive material that may be released from the fuel and the HPB, and the limiting of the potential for air ingress following large HPB breaks. A summary of the alternative concepts is provided in Table 1 and includes two alternatives for a vented and unfiltered building, three alternatives for a filtered and vented building, and two alternatives for a leak tight or pressure retaining reactor building.

In order to gain insight into alternative strategies for implementing the safety functions assigned to the NGNP reactor building, a set of alternative design concepts were developed and evaluated. The evaluations were supported by scoping calculations, including an assessment of the capabilities of each alternative to mitigate the releases from a broad spectrum of LBEs involving a Depressurized Loss of Forced Cooling (DLOFC) condition. These scoping calculations should not be confused with the comprehensive safety analyses that will be required to demonstrate that the actual building design is capable of meeting the functional requirements




in the next phases of the NGNP design. Such comprehensive analyses require design details not currently available and are outside the scope of this study.

Table 1 Alternative Reactor Building Strategies for Performing Safety Functions

RadionuclideFiltration

No. Design Description

VentedAreaLeakRateVol%/day

Pressure Relief Design Features

Postblow-

down re-closureof PRS shaft?

Blow-downphase

Delayedfuel

releasephase

1a. Unfiltered and vented

50-100 Open vent No None Passive

1b Unfiltered and vented with rupture

panels

50-100 Internal + External rupturepanels

No None Passive

2 Partially filtered and vented with rupture

panels


Yes None ActiveHVAC

3a Filtered and vented with rupture panels


Yes Passive ActiveHVAC

3b Filtered and vented with rupture panels

and expansion volume

25-50 Internal + External rupturepanels + expansionvolume


4a Pressure retaining with internal rupture

panels

0.1-1 Internal rupturepanels

N/A Passive Passive

4b Pressure retaining with internal rupture

panels and expansion volume

0.1-1 Internal rupturepanels + expansionvolume

N/A Passive Passive

A comparison of a set of alternatives that were assumed to have the same volume of the Reactor Building vented area, approximately 10,000 m3 with respect to their capability to retain radio-nuclides released from the PHTS pressure boundary for a range of HPB break sizes from 2 mm to 1,000 mm is shown in Figure 1. As seen in this figure the filtered vented reactor building Alternative 3a provides the most effective radionuclide retention capability among the




options considered at this vented area volume including the pressure retaining Alternative 4a.When Alternatives with larger vented area volumes are included, Alternative 3b, which is also a vented filtered configuration, was found to provide superior retention capability among all the alternatives considered. Alternative 4a performs more poorly than 3a or 3b because that option does not remove the pressure driving force that exists following HPB depressurization.Alternatives 3a and 3b on the other hand, despite having much higher leak rate than Alternative 4a or 4b, manage the pressure relief function and eliminate the pressure driving force for the delayed fuel release after the time of depressurization, while using filtration to mitigate the blow-down release and the part of the delayed fuel release that occurs prior to depressurization.

As seen in Figure 1, all of the evaluated alternatives meet the EPA Protective Action Guideline limit and the dose limits for design basis accidents from 10 CFR Part 50.34[8] for design basis breaks up to 100 mm as well as the beyond design basis breaks from 100 mm to 1,000 mm equivalent break size by large margins based on these scoping calculations. The break size of 1,000 mm corresponds to the equivalent single ended break size for a double ended guillotine break of the largest pipe in the PHTS which has an internal diameter of 710 mm for the PBMR NGNP design.

1.0E-03

1.0E-02

1.0E-01

1.0E+00

1.0E+01

1.0E+02

1.0E+03

1.0E+04

1.0E+05

1 10 100 1000

HPB Break Size [mm]

Site

Bou

ndar

y D

ose

TED

E [m

rem

]

Alternative 1a/1bAlternative 2Alternative 3aAlternative 4a

10 CFR 50.34 Limit = 25 rem TEDE

EPA PAG Limit = 1 rem TEDE

Figure 1 Site Boundary Dose (TEDE) vs. HPB Break Size for LBEs Involving DLOFC for Alternative Reactor Building Concepts with Comparable Volumes




Evaluation Summary of Alternative Reactor Building Concepts (Part A)

In general, the vented options that were considered (1a, 1b, 2, 3a, and 3b) were found to be superior to the pressure retaining options (4a and 4b) based on the following considerations:

• Greater compatibility with a non-condensable and inert primary coolant.

• Venting of the primary coolant inventory to atmosphere with or without filtration eliminates a driving force for subsequent fission product transport of the delayed fuelrelease source term.

• When used with filtration (2, 3a, and 3b) provides more effective retention of radio-nuclides for the design basis event spectrum up to 100 mm. Alternatives 3a and 3b provide superior retention for beyond design basis event break sizes up to 1,000 mm as well.

• Lower capital and operating costs.

• Easier and less costly to engineer interfaces with RCCS, SHTS, FHSS, HSS and other NHSS and auxiliary systems.

The highest rating of integrated evaluation for alternatives examined was Alt 2 (Partiallyfiltered and vented with rupture panels) followed by Alt 3a (Fully filter and vented with rupturepanels).

• Both alternatives (2 and 3a) provide superior radionuclide retention capability for design basis HPB breaks with DLOFC than the pressure retaining alternatives (4a and 4b); Alt 3b closely followed by 3a is superior to all evaluated alternatives across the entire HPB break spectrum including AOOs, DBEs, and BDBEs.

• Alts 2, 3a, and 3b are expected to have greater licensability than either of the open vented options (1a and 1b) due to their superior capability to mitigate releases and air ingress.

• Another alternative for future study is a vented building with a passive re-closabledamper without a filter. This is expected to have delayed fuel release retention capabilities approaching that of Alt 2 due to the capability to achieve a lower leak rate.

• Results of the radionuclide retention study show that all the evaluated alternatives provide sufficient margins to offsite dose limits based on inherent and passive safety characteristics of the PBMR NGNP.

• Added engineered features such as filters and re-closable dampers add additional margins.

• This study confirms that radiological retention is not a required safety function but rather a supportive safety function for the NGNP reactor building according to how these terms are defined in the NGNP risk informed and performance based licensing approach.Having said that it is noted that the required safety functions of the reactor building that




involve the structural protection of the reactor and its inherent and passive safety characteristics also serve to maintain the fundamental safety function of controlling radionuclide releases. It supports this function primarily by keeping the radio-nuclidesinside the coated particle fuel and PHTS HPB and secondarily by retaining radio-nuclidesthat may be released from the fuel and HPB.

• In order to support the PBMR NGNP capabilities for defense-in-depth, it is recommended that a design goal be set for a radiological retention capability of a factor of 10 reduction in releases from the RB relative to that released from the HPB for I-131and Cs-137 for DBE and BDBE HPB breaks.

Conclusions regarding radiological retention capability of evaluated RB options are subject to limitations due to:

• Lack of design maturity and associated full scope PRA model• Need to evaluate different HPB break locations and a fuller set of licensing basis events• Need to consider the impact of natural convection on core temperatures during small

leaks (2 to 10 mm)• Lack of a fully integrated mechanistic source term model• Need to consider the failure probabilities of various design features as well as RB

structural capability to withstand loads from a full set of licensing basis events• Lack of a full uncertainty analysis in the source term and consequence modeling

Evaluation of Reactor Embedment (Part B)

This portion considers many factors that influence the selection of embedment depth.There are trade offs to consider between features providing protection from hazards, and design and construction elements that contribute significantly to project cost. The detailed discussion of these factors is included in section B1.

The following is a list of factors that are considered in section B1 and basic finding of the study relative to influence on selection of embedment depth:

• Operational Needs including equipment layoutThe PBMR allows for flexibility in selecting embedment depth. Access for operation and maintenance activities can be accomplished at practically any level. There is no need forroutine access to the reactor from above as there is with other HTGR designs. The pressure relief system and ventilation systems will discharge to atmosphere at the top of the building. There could be some advantage to having a tall building to elevate this release point. At this time it is not expected that an elevated release will be required to meet off site dose limits. A partially embedded building offers an advantage of reduced travel distances for operations and maintenance activities and associated reduction in operating Life Cycle cost. Therefore, it is concluded the partial embedment is favored slightly in meeting operational needs.




• Heat Dissipation to EnvironmentReview of this factor indicates that heat dissipation to the ground or air will not be a design basis mechanism. It is observed, however, that there will be very little difference between heat transfer to soil or to air if this must be considered as last resort. Therefore, it is concluded that this factor has very little, if any, influence on embedment depth selection.

• Water Table EffectsThis factor clearly favors minimum embedment, as dewatering is a cost and potential environmental permitting concern. Hydrostatic pressure causes buoyancy concerns at lower embedment depths. This factor is less important when considering a rock site or soil site with a very deep water table.

• Geotechnical Constraints and Foundation PerformancePartial embedment is favored to achieve improved resistance to overturning moments.Deeper embedment becomes less favorable due the complexity and cost of excavation and foundation design.

• Construction ConsiderationsConstruction is more complex and costly at deeper levels of embedment. Safety challenges will also be more rigorous, in a deeper excavation. This factor favors shallower embedment.

• Cost BenefitIt is shown that costs for excavation, backfilling, dewatering, and external concrete wall construction increase exponentially with embedment depth. This takes into account that robust exterior walls required for hazard protection will be extensive for above ground alternatives, but that below ground level wall and excavation costs will increase dramatically with deep embedment. This concern is more dramatic with a rock site than for a soil site. Therefore this factor favors minimal embedment.

• Malevolent Hazards

Protection from Design Basis Threats (DBT) and Beyond Design Basis Threats BDBT is most easily achieved by full embedment. This is a major concern. However it has been shown on many projects that protective features can be designed for the these hazards,along with natural hazards.

• Natural Phenomenon Hazards

It has been shown on many projects that protective features can be designed for thesehazards. This is a moderate concern that slightly favors full embedment.

• Natural Geological HazardsIt is advantageous to be able to reduce seismic accelerations by partially embedding the structure. This reduction is allowed for up to 50 percent embedment. Below 50 percent




embedment, further reductions are not allowed. Other geological hazards are discussed but do not influence embedment depth significantly. Therefore, this factor favors partial embedment

• Chemical Releases, Explosions, and Manmade HazardsThe hazard imposed by toxic and flammable chemicals on or near the site is dependent on the amount stored and the distance from the reactor building and the control room.Future analysis will be required to ensure that distances and quantities are limited to prevent substantial damage to safety related SSCs. It is expected that the structure will be robust for other reasons. There is a concern that heavier than air gasses could enter a below ground structure. It is concluded that this factor is minor and that it favors a minimally partially embedded structure, only slightly.

Embedment Alternatives Considered for Rock Sites and Soil Sites (Part B)The following alternatives are considered in this study for scoring and ranking purposes.

It is expected that the optimal embedment depth will be developed in conceptual design and will be dependent on specific site conditions and ongoing analysis of operations needs.

• Minimal embedment of the building at approximately 7 to 10 meters

• Partial embedment 20 to 30 meters

• Full Embedment 60+ meters

Scoring Summary for Reactor Embedment (Part B)

The Alternatives for embedment depth as indicated above and discussed in section B2 forrock and soil sites were scored as shown in Table 2. Scoring is conducted for a rock site with deep water table similar the INL site. A soil site with high water table is also scored as it is expected that this will be a very common set of conditions for many attractive commercial sites.A soil site with a low water table is expected to exhibit similar characteristics to a rock site and therefore is not specifically evaluated.

Table 2 Evaluation of Embedment Alternatives for Rock and Soil SitesDesign Alternative Total Weighted

ScoreRemarks

Rock Site Minimum Embedment 510 Maximum exposure to hazards Rock Site Partial Embedment 610 Best Score for rock site with balance

of protection from hazards and reduced cost

Rock Site Full Embedment 465 Water table is not a concern.Excavation and foundation

complexity contributes to construction




Design Alternative Total WeightedScore

Remarks

costSoil Site Minimum Embedment 580 Maximum exposure to hazards but

Benefit with reduced water table concern

Soil Site Partial Embedment 650 Best score for soil site Soil Site Full Embedment 395 Water table and foundation

complexity contribute to very high costs

Recommendations Regarding Embedment (Part B)

The partial embedment scheme scores best for both rock site and soil site, and will therefore be recommended. It offers some protection from hazards without driving costs up excessively as compared to deep embedment. The degree of partial embedment will be optimized for the site during the conceptual design phase.

The PBMR design is quite flexible to accommodate varying site geotechnical conditions.Access to the reactor building can be adjusted for a given site without a significant impact on the layout of major systems within the building.

Role of Reactor Embedment in Satisfying Reactor Building Functions and Requirements (Part B)

The Reactor Building Functions and Requirements portion of this study (Part A) definesreactor building functions and requirements that must be fulfilled independently of the extent of embedment. Part A of this study provides as a preliminary set of licensing basis events involving leaks and breaks in the PHTS and SHTS piping that the building must withstand. PartA also evaluates a number of alternative design strategies for mitigating pipe breaks and minimizing radiological releases from the building. These options can all be applied to any level of reactor embedment. Hence the functions and requirements section is not dependent on the outcome or conclusions of the reactor embedment section.

The reactor embedment section (Part B) is somewhat dependent on the requirementsidentified in Part A. Also, Part B describes external hazards and physical security requirements that the building design must be able to accommodate. The capabilities of the reactor to protect against these hazards can be significantly influenced by the level of embedment.




Basic functional needs to be met by the reactor building that are dependent on level of embedment are to:

• House and provide civil structural support for the reactor and all the SSCs within the scope of the NHSS (building geometry/space)

• Resist all structural loads as required to support both required and supportive safety functions allocated to the NHSB

• Protect the SSCs within the NHSS that perform safety functions from all internal and external hazards as identified in the LBEs as necessary to meet the TLRC.

• Provide physical security of vital areas within the NHSB against acts of sabotage and terrorism

• Provide radionuclide retention during blow-down and delayed fuel release phases of depressurization events

A complete correlation between identified T&FRs and the consideration of reactorembedment is included in Table 31 in section B2.6. This table shows the full list of T&FRs; the role, if any, that embedment serves to accommodate these requirements; and the rational for selecting a particular embedment alternative. T&FRs are also shown in Table 6 in the functions and requirements section.

Open Issues and Additional R&D and Engineering Studies

Lists of open items to be considered in conceptual design and beyond are included in Sections A5 and B3. There are specific R&D items identified. Pending a more thorough design iteration, DDNs will be formulated leading to potential technology development, primarily in the fuel, reactor, and HPB.




INTRODUCTION

This study supports the development of the technical and functional requirements (T&FRs) for the NGNP reactor building, including consideration of requirements forembedment. The approach to the definition of these requirements includes consideration of:

• NGNP top level requirements

• Role of Reactor Building in the PBMR NGNP safety design approach

• NRC regulations and guidance relevant to the Reactor Building

• NRC regulations on design basis threats and hazards

• Definition of Reactor Building safety functions including those that are required to meet requirements for mitigating design basis accidents as well as those that support the prevention and mitigation of a wide spectrum of licensing basis events

This study shall be based on the current understanding of expected source terms for the NGNP licensing basis events, including a comprehensive set of fission product transport phenomena from the fuel into the PHTS circuit, from the PHTS circuit into the Reactor Building, and from the Reactor Building to the environment. As such, it is understood that this will be a scoping study helping to frame the issues associated with development of the T&FRs for the Reactor Building. Accordingly, an objective of this study is to identify the issues and further R&D and engineering studies that are required to resolve these issues.

This study also develops the requirements and criteria for determining the degree of embedment of the reactor building. This includes embedment studies for the NGNP reactor building concepts, considering the interaction among factors that influence the depth of the embedment. These factors include cost, design basis threats, seismic effects, hazards resistance, etc. The results of this study will be used to characterize the interactions of these factors on embedment depths for commercial application of this technology.




OBJECTIVE AND SCOPE

The objective of this study is to develop the technical and functional requirements(T&FRs) for the NGNP and HTGR reactor building, including consideration of reactorembedment. The scope of the study includes Part A, in which the overall technical and functional requirements for the reactor building are defined and various design strategies to meeting these requirements are evaluated. In Part B an evaluation is made of strategies involving reactor embedment.

This study is organized into two parts based on the assumption that the type of reactor building can be separated to a certain extent from the degree of reactor embedment. However,these parts are organized into a single report in view of the expected synergy between the tasks of identifying a comprehensive set of reactor building requirements and the evaluation of reactor building design options, and the reactor embedment options against these requirements.

The development of T&FRs for the reactor building will be based on a review of:

• NGNP user requirements associated with safety, availability, investment protection, plant costs, and the commercialization business case.

• NRC regulations and HTGR objectives regarding the frequency and radiologicalconsequences of licensing basis events, SSC safety classification, and defense-in-depth.

• NRC policy issues (SECYs) and staff requirements memoranda (SRM) on the expected safety characteristics of advanced reactors and the development of containment performance criteria for HTGRs.

• NRC regulations on design basis and beyond design basis threats and hazards to be addressed for existing and advanced reactors.

• Requirements for retention of radio-nuclides within each barrier including the fuel, primary helium pressure boundary, and the reactor building, and, if needed, definition of reactor building release rate and filtration requirements.

This study is based on the current understanding of expected categories of LicensingBasis Events (LBEs) as well as an expected range of frequencies and mechanistic source termsfor these LBEs that are appropriate for the PBMR. As such, it is understood that this is a scopingstudy helping to frame the issues associated with development of the T&FRs for the reactor building. Accordingly, an objective of this study is to identify the issues and further R&D and engineering studies that are required to resolve these issues. This study investigates alternative design strategies to meet the reactor building technical and functional requirements including the use of various design approaches to address the required and supportive safety functions of the reactor building. The required safety functions are focused on the preservation of the reactor




core geometry and structural protection of the reactor and safety classified SSCs from a range of internal and external hazards and challenges represented in the LBEs. Retention of radio-nuclides that are released from the fuel and the PHTS HPB is a supportive safety function for the reactor building because the primary safety function of retaining radio-nuclides is assigned to the fuel in the NGNP safety design approach

This study includes embedment studies for the PBMR RB concepts, considering the interaction among factors that influence the depth of the embedment. These factors include cost, design basis threats, seismic effects, hazards resistance, etc. The results of this study are used to characterize the interactions of these factors on embedment depths for commercial application of this technology. The recommendations from relevant sections of the Electric Power Research Institute (EPRI)’s Advanced Light Water Reactor Utility Requirements Document [24] areevaluated for applicability in this study. This phase of the study also includes a review of prior NRC reviews of modular HTGR designs and the conclusions from those reviews concerning the embedment feature of the technology on licensing considerations.

ORGANIZATION OF REPORTThis study is organized into two parts based on the assumption that the type of reactor

building can be separated to a certain extent from the degree of reactor embedment. Howeverthese parts are organized into a single report in view of the synergy between the tasks of identifying a comprehensive set of reactor building requirements and the evaluation of reactor embedment options against these requirements.

Part A of this report covers the development of T&FRs for the reactor building and the identification and evaluation of various design strategies for meeting these requirements. These strategies include those to perform required safety functions such as pressure relief from HPB depressurization events as well as those to perform the supportive safety function of retention of radioactive material releases from the HPB and limiting of air ingress. All the strategies evaluated in Part A assume a nominal extent of reactor embedment consistent with the design of the PBMR Demonstration Power Plant. Part B addresses that part of the scope of work dealing with reactor embedment, whose range of options vary with the site, from fully above to fully below grade.




PART A: REACTOR BUILDING TECHNICAL AND FUNCTIONAL REQUIREMENTS AND EVALUATION OF

ALTERNATIVE CONFINEMENT STRATEGIES

The purpose of Part A is to define the technical and functional requirements for thereactor building and to evaluate different design concepts toward fulfilling these requirementsindependent of embedment, which is evaluated in Part B. In Section A1 the derivation of thetechnical and functional requirements for the Reactor Building is documented. An evaluation of alternative design concepts for implementing these requirements is discussed in Section A2. InSection A3 the results, conclusions, and recommendations are summarized.

A1 DEFINITION OF TECHNICAL AND FUNCTIONAL REQUIREMENTS

A1.1 TECHNICAL APPROACH TO DEFINITION OF REQUIREMENTS

The technical and functional requirements (T&FR) for the NGNP Reactor Building are comprised of a list of reactor building functions and a specification of the technical requirements for satisfying these functions. The technical approach to defining these functions is outlined in Figure 2 and consists of the following steps implemented in a top-down fashion:

• Review the NGNP Top Level Requirements in the PCDR (Reference [1])

• Review NRC requirements applicable to the Reactor Building

• Identify the role of the Reactor Building in the PBMR NGNP safety design approach (References [2] through [6])

• Consideration of the approach to defining safety functions in the risk-informed and performance based approach envisioned for the PBMR NGNP

• Development of a preliminary list of licensing basis events for the Reactor Building

• Definition of required and supportive safety functions for the Reactor Building and associated technical requirements

• Definition of physical security requirements for malevolent human actions

• Definition of non-safety related requirements for the Reactor Building

• Compilation of a complete list of Reactor Building technical and functional requirements

The results of each step in this process are presented in the sections below.




Figure 2 Flow Chart of Part A: Development of Reactor Building Requirements and Evaluation of Alternative Concepts




A1.2 NGNP TOP LEVEL REQUIREMENTS

The NGNP top level requirements for the reactor building and the rest of the plant flow from the Top Level Requirements set forth in of the NGNP Preconceptual Design Report(PCDR), Section 2 [1]. These Top Level Requirements flow down to requirements impacting the reactor building and include:

• Plant performance requirements

• Plant availability and investment protection requirements

• Plant cost targets that are in alignment with the NGNP commercialization strategy

• Plant safety and associated licensing requirements

The NGNP safety and associated licensing requirements include the overall NGNP requirements as well as specific requirements for the Nuclear Heat Supply System (NHSS) and other NGNP facilities. Five safety and licensing design basis requirements are listed for the NHSS:

• The NHSS shall be designed in accordance with the generic top-level, nuclear regulatory criteria that are direct and quantitative measures of nuclear-related risks and consequences, plus all applicable governmental codes and regulations.

• The NHSS shall meet the top-level regulatory criteria without credit for sheltering or evacuating the public beyond the plant's exclusion area boundary, with the intended result that the Emergency Planning Zone (EPZ) is limited to the Exclusion Area Boundary (EAB).

• The NHSS shall meet the top level regulatory criteria without reliance on prompt operator action, the control room or its contents, including the automated plant process control system, or on auxiliary power supplies (other than batteries).

• Design requirements supporting safety and licensing objectives, that are incremental to those employed in conventional power plant design, construction, operation and maintenance and quality assurance practices, shall be developed using the principles of risk-informed, performance-based regulation.

• The NHSS shall be equipped with provisions to safely shutdown the reactor in the unlikely event the central control room becomes uninhabitable.

Three additional safety and licensing requirements are listed for the Hydrogen Production Facility (HPF):

• The hydrogen production facilities, including the conversion, storage, and distribution systems, shall comply with the requirements of 29 CFR 1910.103, Occupational Safety and Health Standards, Subpart H - Hazardous Materials, Hydrogen [51].




• In the event that the HPU facility also produces and stores significant quantities of oxygen, the requirements of 29 CFR 1910.104, Oxygen [52], shall be applied.

• The design, operation and maintenance of the HPU shall comply with 29CFR 1910.119, “Process Safety Management of Highly Hazardous Chemicals” [53].

The above listed requirements shall be used as a starting point for formulating reactor building-specific technical and functional requirements.

A1.3 REVIEW OF APPLICABLE NRC REQUIREMENTS

Existing regulatory requirements for reactor buildings and containment of radioactive material are specific to LWRs and reflect the safety design approach of LWRs. That approach involves the use of active engineered safety systems to prevent core damage over a spectrum of design basis events with a focus on loss of coolant accidents and a pressure retaining containment structure to mitigate the consequences of design basis and to contain radioactive material that could be released in beyond design basis core damage accidents. The PBMR NGNP is based on a fundamentally different safety design approach that identifies a different set of safety requirements for the barriers to radionuclide release, including the reactor building, as discussed more fully in the next section. As a result of these differences, the existing requirements for reactor buildings and containment of radioactive material need to be interpreted with great care. The regulatory requirements for the NGNP Reactor Building will need to be resolved in the course of implementing the RI-PB approach.

Table 3 Elements of NGNP Risk-Informed Performance-Based Licensing Approach

Elements of Approach Purpose

• Top Level Regulatory Criteria (TLRC) • Establish what level of safety must be achieved in terms of the frequencies and radiological doses of event sequences

• Licensing Basis Events (LBEs) • Define when and under what conditions the TLRC must be met based on event sequences selected from the PRA

• Required Reactor Specific Safety Functions

• Regulatory Design Criteria (RDC)• Safety Classification of SSCs• Plant Capability Defense-in-Depth

• Establish how it will be assured that the TLRC are met




Elements of Approach Purpose

• Design Basis Accident (DBA) Conditions• Special Treatment Requirements• Programmatic Defense-in-Depth

• Provide assurance as to how well the TLRC are met by satisfying deterministic requirements to reduce uncertainties in the probabilisticresults

The NGNP PCDR specifies a risk informed and performance based (RI-PB) licensing approach to derive regulatory requirements for the NGNP. The key elements of the RI-PBlicensing framework are summarized in Table 3.

NRC Research has developed a similar RI-PB licensing approach for advanced non-LWRs referred to as the Technology Neutral Framework as described in NUREG-1860 [7]. A review of this document provides an indication of the NRC research staff current thinking on how to develop a set of regulatory requirements for an advanced non-LWR. The NUREG-1860[7] method is often referred to as the NRC Technology Neutral Framework (TNF). A comparison of the NGNP RI-PB approach as described in Section 14 of the PCDR and the TNF indicates a number of comparable features including the following.

Both approaches:

• Use frequency vs. dose criteria derived from regulations to judge the acceptability of licensing basis events (LBEs)

• Derive LBEs from a full scope all modes PRA• Assign LBEs to three categories based on frequency of the underlying event sequences• Include conservative deterministic analysis to augment the PRA• Derive special treatment requirements for SSCs based on the LBE results and frequency-

dose criteria• Apply defense-in-depth criteria to address uncertainties• Include a need to screen the existing requirements for applicability and to develop new

technology specific requirements

Some of the key differences between the respective approaches are listed in Table 4.

Appendix H of NUREG-1860 [7] provides NRC staff screening of regulations for applicability to new non-LWRs. The process that was followed and the results of the screeningappear to be similar to screening of regulations performed by Exelon during earlier pre-licensinginteractions associated with the PBMR. The scope of this screening includes all of 10 CFR and each of the GDCs in Appendix A to 10 CFR Part 50 [54]. Most GDCs are either retained as currently written or revised to make the criteria reactor technology neutral (54 GDCs fall into this category). Some GDCs were assessed by the staff to be specific to LWRs and not applicableto non-LWRs (GDC 33, 35-40, 44-46). It is noteworthy that GDC-3 on fire protection is revised in this research report revised to include graphite fires.




The following conclusions and insights for the NGNP Reactor Building were developed as a result of this review of NRC regulatory requirements:

• The current NRC requirements for reactor buildings and containment of radioactive material are specific to LWRs and need to be interpreted for applicability to the PBMR NGNP.

• Formulation of NGNP Reactor Building requirements requires some assumptions about the outcomes of applying and NRC acceptance of the NGNP RI-PB approach

• Similarities with NUREG-1860 [7] approach encouraging as to the acceptability of the NGNP RI-PB approach

• NRC staff appears to expect a radiological confinement function for the NHSB and has concerns about graphite fires which will need to be addressed.

• In principle, it appears the staff is prepared to accept a non-pressure retaining vented confinement with some controls on leak rate and significant radiological retention capability within the NHSB and associated systems.

• A key technical issue to be resolve is getting agreement with the NRC on the selection of licensing basis events for evaluating the capability of the Reactor Building to satisfy its safety functions and associated licensing requirements.

• Key issues will be acceptance of NGNP PRA, mechanistic source terms for LBEs, and reasonable resolution of the so-called “deterministic” design basis event for evaluation of the third barrier and siting.

Table 4 Key Differences between NGNP and NUREG-1860 RI-PB Approaches

Characteristic NGNP RI-PB Approach NUREG-1860 RI-PB Approach

Frequency metric Events per plant-year Events per reactor-year

AOO; = 10-2/plant-year Frequent: 10-2/Rx-year

DBE; 10-4 to 10-2/plant-year Infrequent: 10-5 to 10-2/Rx-year

LBE Categories

BDBE: 5x10-7 to 10-4/plant-year Rare: 10-7 to 10-5/Rx-year

Safety related (SR) Safety significant (has special treatment)

Non-safety related with special Treatment (NSRST)

SSC Safety Categories

Non-safety related without special treatment

Non-Safety Significant (no special treatment




Characteristic NGNP RI-PB Approach NUREG-1860 RI-PB Approach

DeterministicSafety Analysis

Deterministic DBAs must meet design basis dose limits with mitigation credit for only SR

SSCs and conservative assumptions

Frequency vs. dose criteria must be met with mitigation credit for

only safety significant SSCs

Specific additional deterministic requirements for frequent and infrequent events; Additional

deterministically selected LBE

A1.4 DEFINITION OF SAFETY AND LICENSING REQUIREMENTS

A1.4.1 Role of Reactor Building in NGNP Safety Design Approach

In order to provide a context for defining the safety and licensing issues that influence the reactor building design it is necessary to review the NGNP safety design approach and the role of the reactor building in prevention and mitigation of licensing basis events.

The risk-informed and performance-based safety design approach is derived from that developed by PBMR (Pty) Ltd. in support of the design certification for future U.S.-sited plants.This PBMR approach builds upon the successful application of risk-informed methods undertaken as backfits for the current fleet of licensed, operating reactors and is an extension to the methods used in other design certification applications. (References [3], [4], [5] and [6])

The safety design philosophy is to apply the principles of defense-in-depth at a fundamental level in which a diverse combination of inherent reactor characteristics, passive design features and Structures, Systems and Components (SSCs), active engineered systems, and operator actions are deployed to maintain the integrity of robust passive barriers to radionuclide release. These barriers include the fuel barrier, the primary helium pressure boundary, and the reactor building. The plant capabilities that support defense-in-depth for any reactor technology are illustrated conceptually in Figure 3. For any reactor, these capabilities are based on a foundation of inherent reactor characteristics, a set of barriers to the release of the reactor’s inventory radioactive material, and a set of passive and active SSCs that perform safety functions that can be tied to the preservation of the integrity of one or more barriers. For different reactor concepts, the inherent characteristics are different, and reactor safety design approach for use of combinations of passive and active design features is also necessarily different.




Fuel Barrier

Coolant Pressure Boundary

Reactor Building Barrier

Core InventoryOf Radioactive

Material

PassiveEngineered

Features and SSCs

ActiveEngineered

Features and SSCs

Inherent Properties of Core, Fuel, Moderator, Coolant, and Reactor Vessel

Safety FunctionsProtecting Barriers

Site Selection

Fuel Barrier

Coolant Pressure Boundary

Reactor Building Barrier

Core InventoryOf Radioactive

Material

PassiveEngineered

Features and SSCs

ActiveEngineered

Features and SSCs

Inherent Properties of Core, Fuel, Moderator, Coolant, and Reactor Vessel

Safety FunctionsProtecting Barriers

Site Selection

Figure 3 Key Elements of Plant Capabilities Supporting Defense-in-Depth

All reactors must fulfill a set of fundamental safety functions that are derived from a need to prevent the release of radioactive material and to mitigate the extent of releases during accidents. The specific set of safety functions and the allocation of passive and active SSCs to perform these safety functions are necessarily specific to each reactor and dependent on its safety design philosophy. The safety design approach allocates responsibilities to each barrier for the prevention and mitigation of accidents which are defined in terms of the events and event sequences that challenge the safety case.

The NGNP-specific key safety functions are derived in a top-down manner with the objective of protecting the integrity of the multiple barriers to radionuclide release. A fundamental aspect of the safety design philosophy is to provide the capability to perform safety functions first through the selection of inherent reactor characteristics and engineered systems that operate on passive design principles and then to support these safety functions withcombinations of diverse active engineered systems and operator actions.

PBMR safety philosophy starts with the application of defense-in-depth principles in making fundamental design selections:

• Inherent reactor characteristics

• Multiple, robust, and concentric barriers to radionuclide release

• Conservative design approaches to support stable plant operation with large margins to safety limits




• In the event safety functions are challenged, the safety philosophy is to utilize the inherent reactor characteristics and passive design features to fulfill the required safety functions

• Additional active engineered systems and operator actions are provided to reduce the challenge to plant safety and provide defense-in-depth in preventing and mitigating accidents

• Avoid need for early operator intervention, or early functioning of any active systems to maintain safe stable state

The NGNP safety design approach is framed in terms of reactor-specific safety functions that were developed from the top goal of retaining the inventory of radio-nuclides primarily within the fuel and then considering the specific functions that, when satisfied, would protect the integrity of the fuel and other radionuclide transport barriers. These safety functions, which are illustrated in Figure 4, are comprised of two general categories referred to as required and supportive safety functions. These categories are defined as follows:

• Required Safety Functions—Those functions that are necessary and sufficient to meet the dose limits for Design Basis Events and deterministically selected Design Basis Accidents.

• Supportive Safety Functions—All other functions that contribute to the prevention or mitigation of accidents and support the plant capabilities for defense-in-depth.

Maintain Control of Radionuclide Exposure

ControlRadiation

Control Personnel Access

Control Radiationfrom Processes

Control Radiation from Storage

Control Radiationfrom Core

Control Radiation Transport

Control Direct Radiation

Control Transport from Site

Control Transport in Reactor Building

Control Transport from HPB

Control Transport from Core

Retain Radionuclides in Fuel Elements

Retain Radionuclides in Fuel Particles

Remove Core Heat Control Core Heat Generation

Control Chemical Attack

Required Safety Function

Supportive Safety Function

Figure 4 Top Down Development of NGNP Safety Functions




In the future NGNP licensing submittal, analyses will be included to demonstrate which safety functions are classified as required and which as supportive according to the NGNP RI-PBas discussed more fully in Section 14 of the PCDR (Reference [2]). The required safety functions for the NGNP, all of which contribute to the fundamental safety function of preventing and mitigating the release of radioactive material, include those functions to:

• Control heat generation

• Control heat removal

• Control chemical attack

• Maintain core and reactor vessel geometry

• Maintain reactor building structural integrity

In formulating requirements for the Reactor Building, the following types of functions need to be considered:

• Safety Functions o Required Safety Functions assigned to the reactor building

o Supportive Safety Functions assigned to the reactor building

• Physical Security Functions—Those functions assigned to the building protect the reactor and vital equipment from design basis threats associated with acts of sabotage and terrorism

• Other Functions—Functions necessary for plant construction, operation, maintenance, access, inspection, worker protection, and control of routine releases of radioactive material during normal operation

The scope of requirements and technical functions for the reactor building are allocated in the PCDR to the Nuclear Heat Supply Building (NHSB), NHSS HVAC System, and NHSB Pressure Relief System (PRS). Based on the NGNP safety design philosophy and the NGNP RI-PB licensing approach, the following required and supportive safety functions can be defined:

The NHSB shall perform the following required functions:

• House and provide civil structural support for the reactor and all the SSCs within the scope of the NHSS

• Resist all structural loads as defined for NGNP required safety functions allocated to the NHSS

• Protect the SSCs within the NHSS that perform required safety functions from all internal and external hazards as identified in the LBEs




• Provide physical security of vital areas of the NHSS against acts of sabotage and terrorism

The RB shall perform the following supportive safety functions:

• Provide retention of radio-nuclides released from the PHTS HPB

• Limiting the potential for air ingress into the PHTS following HPB leaks and breaks as defined in the licensing basis events

Note that the final list of SSCs that are classified as safety related and therefore must be protected by the building from DBE loads is to be determined in future stages of design and is outside the scope of this study. In reviewing the above allocation of safety functions to required and supportive categories, it must be kept in mind that all the NGNP safety functions support the fundamental safety function of preventing and mitigating the release of radioactive material.According to the NGNP safety design approach, the allocation of safety functions to the reactor building is of a structural nature. If the NGNP Reactor Building fulfills its required safety function of structurally protecting the core geometry and the structural integrity of the PHTS pressure boundary, it will also be supporting the required safety functions of controlling core heat removal, preventing chemical attack and controlling heat generation. These in turn serve to keep the radionuclide inventory inside the coated fuel particles and when radioactive material is released from the fuel particles during normal operation or during an accident, helps to keep the releases from the PHTS HPB and Reactor Building within their required limits. Having met those required safety functions, the functions of retaining radionuclide that may be released from the PHTS and those of limiting the potential for air ingress are classified as supportive because they are not required or credited in the performance of the deterministic design basis accidents.SSCs that perform required as well as supportive requirements are subjected to special treatment requirements to the extent necessary to keep the frequencies and consequences of the licensing basis events within the TLRC, according to the NGNP RI-PB licensing approach as discussed more fully in Section 14 of the PCDR (Reference [2]).

It is noted that the Reactor Cavity Cooling System (RCCS) also provides important functions that contribute to the structural integrity to the reactor vessel and the maintenance of core geometry. Those functions are primarily supported during normal plant operation prior to the initiating event by keeping the vessel support temperatures and reactor cavity concrete temperatures within code limits. In the event that continued operation of the RCCS following any design basis event is deemed to be required to maintain structural integrity, a required safety function of the reactor building will be to protect the RCCS during all the design basis events.However, further study will be needed to establish whether there is a need for such a requirement.




A1.4.2 Implications of RI-PB Licensing ApproachIn order to obtain a license to operate the NGNP, the reactor building design and the

remaining safety features of the plant will ultimately be subjected to a rigorous deterministic and probabilistic safety analysis guided by a risk-informed and performance-based licensing approach that is described in References [3] through [6]. The safety evaluation for the NGNP will help to define and finalize reactor building requirements and to demonstrate the building design is capable of meeting these requirements. The key elements of this RI-PB approach include:

1. The use of accident frequency vs. radiological dose criteria that are derived from current U.S. licensing requirements, referred to as Top Level Regulatory Criteria (TLRC). These regulatory criteria are expressed in terms of acceptable combinations of event sequence frequencies and radiological doses to the public at or beyond the site boundary as well as the aggregate public health and safety risk due to accidents for demonstrating conformance to the NRC Quantitative Safety Objectives and associated safety goals.

2. Use of a full-scope Probabilistic Risk Assessment (PRA) to select the Licensing Basis Events (LBEs). Full scope means that the PRA will cover all operating and shutdown modes, all internal and external plant hazards, mechanistic source terms and offsite radiological consequences. These LBEs include: Anticipated Operational Occurrences (AOOs) whose frequencies are above 10-2 per plant year and are evaluated using realistic assumptions; Design Basis Events (DBEs); whose frequencies are between 10-4 and 10-2,per plant year and are evaluated using realistic and conservative assumptions, the probabilistic and deterministic safety analysis, respectively, and Beyond Design Basis Events (BDBEs) with frequencies less than 10-4 and whose consequences are evaluated using realistic assumptions. These rules for conservative and realistic assumptions will impact the definition of reactor building performance requirements for different LBEs.

3. Development of reactor-specific functions, selection of the corresponding safety-relatedSSCs, and their regulatory design criteria. The SSC of interest in this evaluation is the reactor building.

4. Deterministic design conditions and special treatment requirements for the safety-relatedSSCs including certain aspects of the reactor building design.

5. A risk-informed evaluation of defense-in-depth.

In this evaluation, a qualitative review of this RI-PB approach was be used to help formulate the preliminary set of reactor building requirements. Insights from existing and previous PBMR and HTGR PRAs and a review of the NGNP design features are used in the next section to identify a set of LBEs for use in this evaluation, in lieu of actually performing a PRA.However, the performance of a full scope PRA and the execution of each of the above steps in the safety evaluation approach will need to be performed to finalize the Reactor Building requirements and to demonstrate the adequacy of the safety case.




A1.4.3 Preliminary Licensing Basis Events

The Licensing Basis Events (LBEs) for the NGNP will be defined via full scope PRA during conceptual and preliminary design. For the purposes of this study, LBEs are defined by engineering judgment based on reviews of the plant design as describe in the PCDR and experience by the authors in the performance of HTGR and specifically PBMR PRAs.

Categories of LBEs include:

• Internal Eventso HPB leaks and breaks in PHTS and SHTS HPBo PHTS HPB Heat exchanger failures (e.g. CCS Hx, IHX)o Transients with and without reactivity addition

• Internal Plant Hazard Eventso Internal fires and floodso Rotating machinery missileso High energy line and tank breaks

• External Plant Hazard Events o Hydrogen Process eventso Seismic Eventso Aircraft crashes and transportation accidentso High winds and wind generated missiles

For each of the above categories the LBEs may include events in either of the three LBEcategories (i.e., AOOs, DBEs, or BDBEs). In the formulation of Reactor Building requirements,a key consideration is the definition of LBEs associated with leaks and breaks on the PHTS and SHTS helium pressure boundary (HPB). This is key consideration because leaks and breaks result in depressurization of large inventories of pressurized Helium and relatively high temperature and resulting pressure and temperature loads on the building and the NHSS SSCs contained within it. Importantly, leaks and breaks in the PHTS HPB also involve a breach in one of the three radionuclide transport barriers, which means that radioactive material released from the fuel, either during normal operation or during an accident, has the potential for release into the building – and if not confined within the building, to the environment. Based on a qualitative review of the NGNP design and a comparison with other designs for which HPB leak and break initiating event frequencies have been quantified, the following categories of HPB leak and Break events were defined for the reactor building. The leaks and breaks in the PHTS are also summarized in Table 5. Aircraft crashes may include those within and outside the design basis depending on the frequency of occurrence.




Table 5 Preliminary LBEs Involving HPB Leaks and Breaks

LBE Type Frequency Range (per plant year) Break Category Equivalent Break

Size (mm)AnticipatedOperationalOccurrences

= 10-2Small PHTS HPB break with forced

coolingSmall PHTS HPB break with loss of

forced cooling

1 to 10

Medium PHTS HPB break with loss of

forced cooling> 10 to 100Design Basis Events 10-4 to 10-2

Large SHTS break with intact PHTS HPB

Beyond Design Basis Events <10-4

Large PHTS HPB break with loss of

forced cooling

> 100 to 1,000*

* 1,000 mm is the equivalent single ended break size for a double ended guillotine break of the largest pipe on the PHTS pressure boundary with inside diameter of 710 mm.

• Anticipated Operational Occurrences (AOOs) assumed to include leaks up to 10 mm in break size anywhere along the PHTS and SHTS HPB inside the RB–PRS and HVAC designed to keep HVAC running during such leaks

• Design Basis Events (DBEs) are expected include:o Up to 50 mm breaks in the FHSS pipes above or below the RPVo Up to 100 mm breaks in the CIP, COP or other PHTS piping above or

below the core including pipe to vessel nozzle weldso Breaks in HSS piping bounded by 100 mm break sizeo Up to 1 meter (SEGB) breaks in the SHTS piping inside the NHSBo Requirement for DBEs is to maintain structural integrity and avoid

damage to any safety related SSC inside the NHSB• Beyond Design Basis Events (BDBEs) are expected to include

o Up to 1 meter (SEGB) breaks in the PHTSo Requirement is to meet TLRC using realistic assumptions and to show that

structural integrity of the RPV and major PHTS SSCs is maintained

In the above for BDBEs the term realistic is used to describe best estimate evaluations that are technically justified and supported by an evaluation of the underlying uncertainties.




A1.5 DEFINITION OF PHYSICAL SECURITY REQUIREMENTS

The design of the reactor building should include consideration of requirements imposed by the need for physical security protection of the plant from acts of terrorism and sabotage. The basis for these requirements and for satisfying them comes from Rules and guidance of the NRC and industry standard responses to these Rules and guidance. The overall goal of physical security is to protect against radiological sabotage. This goal has been further defined as having a physical design and trained response force available to deny access of the “Design Basis Threat (DBT)” from target sets of “Vital Equipment.”

The Design Basis Threat (DBT) is imposed by Title 10 of the Code of Federal Regulations. It is defined by NRC and classified as Safeguards Information (SGI). The DBT defines sets of attackers with definitions of their capabilities, skills and equipment. It defines the number and type of personnel within the DBT, both from inside and outside the facility. It defines the transportation equipment, weapons, explosives and other equipment that are part of the DBT. It defines allowable deployment of the DBT for evaluation.

Security for the reactor building is governed by Parts 50, 52 and 73 of Title 10 of the Code of Federal Regulations [54],[55] and [56]. Security, as defined by 10CFR73 [56] is the protection from radiological sabotage or protection from theft and diversion of special nuclear materials. It must be provided throughout the life cycle of nuclear power plant fuel. Additional,more detailed requirements are in 10CFR73 [56]:

• Vital area means any area which contains vital equipment.

• Vital equipment means any equipment, system, device, or material, the failure, destruction, or release of which could directly or indirectly endanger the public health or safety by exposure to radiation. Equipment or systems which would be required to function to protect public health and safety following such failure, destruction or release are also considered vital.

• [V]ital equipment [shall be located] only within a vital area, which in turn, shall be located within a protected area such that access to vital equipment requires passage through at least two physical barriers of sufficient strength to meet the performance requirements of . . . this section. More than one vital area may be located within a single protected area.

• The physical barriers at the perimeter of the protected area shall be separated from any other barrier designated as a physical barrier for a vital area or material access area withinthe protected area.

• Unescorted access to vital areas . . . shall be limited to individuals who are authorized access to the material and equipment in such areas and who require such access to perform their duties




The designer must define a set of Vital Equipment. Vital Equipment and other spaces (the Control Room, Central Alarm Station and Secondary Alarm Station) must be included in Vital Areas. There must be two barriers between the general public and entry into a Vital Area.Usually one of these barriers is the site Protected Area Boundary, which includes a Vehicle Barrier System. For the Reactor Building, this means that it must support controlled access into Vital Areas and DBT denial from Vital Equipment.

The final design deliverable for physical design basis security is a Physical Security Plan (PSP) that is integrated with the physical design of the facility. The PSP is submitted by the plant owner/operator/licensee to the NRC. It is SGI and is a completion of the Security Plan template developed in cooperation between the nuclear industry (NEI) and the NRC. This template is NEI 03-12 [25] and is the basis for all Physical Security Plans used in operating plants and those currently obtaining licenses. It is assumed that the reactor building, its fuel handing facility and the hydrogen generation facility are collocated, but are such that they can have unique access controls.

The current working definition of “directly or indirectly endanger the public health or safety by exposure to radiation” includes releases from the site in excess of 10CFR100 [57]allowables, radiological sabotage and theft or diversion of formula quantities of special nuclear materials. Documented evidence of adequate protection is required during the Combined License application process. It must include description of design features to deny or delay unauthorized access of the Design Basis Threat (DBT) and the resistance to and mitigation of damage caused by beyond DBT. The configuration of perimeter fencing and intrusion detection will be developed by the combined License applicant. Denial and delay features of the reactor building must be included at the Design Certification phase of licensing. Release or theft of fuel spheres, especially irradiated spheres, could directly or indirectly endanger the public health or safety by exposure to radiation. This release of spheres could also lead to radiological sabotage.

The plant design and defensive strategy will be reviewed against the DBT. This review will include threat definitions, attack scenarios, target set definitions, defensive positions, delay features and access paths. Target sets are those sets of systems, structures or equipment that if compromised or used improperly will lead to radiological sabotage or theft or diversion of special nuclear materials. This means that target sets will be identified within the reactor building. Fault tree analysis and probabilistic review of potential target sets is conducted to determine if they need to be considered. The acceptance criterion is that no member of the DBT disables any target set to the point of release or sabotage.

The review is conducted by security professionals with a table top exercise. This effort is a virtual exercise in which the DBT is deployed against the plant security force and design features. A variety of assaults are conceived to ensure that there are no weak spots in the defensive strategy. This security assessment will use reasonable response parameters to determine the success of the design’s defensive strategy, including the physical design’s delay, denial and mitigation features. From these table-top exercises on other new reactor designs, as




well as live force-on-force exercises at operating plants, many general guidelines have been developed. Some are delineated below.

General practice has a number of barriers between the general public and vital equipment.From the outside in, they are: the Owner Controlled Area boundary, the Protected Area (PA) boundary, the delay fence, and the Vital Area (VA) boundary. Each has a level of increasing access control requirements. As each barrier is crossed, different access permission is required and personnel searching is performed. Each barrier will have defined access ports for both pedestrian and vehicular traffic.

Of these barriers, the reactor building will be a part of only the VA boundary. The VAboundary is a limited access, hardened boundary that forms the final barrier to vital equipment or areas. The exact portions of the reactor building that forms part of the VA boundary will be determined after preliminary determination of target sets. Access control will include automaticand security officer features. The reactor building portions of the VA boundary must include security officer protection and response features. The VA boundary must be resistant to DBT blast definitions.

Rules of thumb and guidelines that, if followed, will make the nuclear plant more secure are:

1. Define all Vital Areas

2. Limit the number of access paths into the facility and the VA, consistent with life safetyrequirements

3. Harden the VA boundary consistent with the DBT and table top results

4. Provide for protection of response personnel

5. Provide for natural “funneling” of approaching attackers

6. Provide for access controls specific to the defense level of the area

7. Make the facility resistant to a large fire or explosion

8. Have clear fields of fire from areas inside the VA to areas outside the VA

9. Ensure that adequate spent fuel cooling is available following a beyond design basisevent

10. Establish a preliminary guard force that meets requirements of alarm station monitoringand threat response

A1.6 REACTOR BUILDING REQUIREMENTS SUMMARY

This study uses the NGNP Pre-Conceptual-Design Report (PCDR) reactor building design as a starting point [1]. It is a rectilinear building, configured to house a 500 MWth PBMR which is coupled to an intermediate cooling loop via two in-series intermediate heat exchangers




(IHXs). The primary loop flow is provided by an electric-drive helium circulator. Thermal energy is transferred in the IHXs to a secondary helium loop, and then on to the power conversion and hydrogen production processes. Figure 5 shows a cut-away view of the PCDR reactor building design and its separation with adjacent buildings.

ReactorBuildingReactorBuilding

Figure 5 NGNP Reactor Building in Preconceptual Design Report

The major components and systems located within the PDCR reactor building include the following:

• Reactor Unit

• Core Conditioning System

• Reactor Cavity Cooling System

• Fuel Handling and Storage System

• Helium Services System

• Heating, Ventilating, and Air Conditioning Systems

• Pressure Relief System

• Other NHSS Support Systemso Portions of the Heat Transport System




o Primary Heat Transport System

o IHX vessels, circulators

o Parts of the Secondary Heat Transport System.

The top-level goals for the NGNP Project include the design, construction, and demonstration of a high-temperature gas-cooled nuclear reactor that will exhibit the capability of both electrical and hydrogen production. In addition, the project will demonstrate features and attributes that are unique to HTGRs, including passive safety and defense-in-depth that relies primarily on inherent design capability and minimal operator action.

This study is based on the use of a pebble-bed modular reactor (PBMR) derived from development work on-going in South Africa, and is focused on three key aspects of the NGNP reactor building:

• Protection of the Reactor Building structure and safety related SSCs contained therein from the effects of pressure and temperature loads from events involving Heat Transport System pipe breaks and depressurization

• Mitigation of radiological consequences following a postulated accident event• The extent to which the NGNP reactor building should be embedded

The need for retention of radio-nuclides by the reactor building will dictate building features such as the boundary leak rate, the need for barriers to internal flow communication, and the need for engineered features like filters at the building exhaust point. These features are investigated in Section A2. The extent to which the building is set below grade will impact, and be impacted by, the need for radionuclide retention, so it is appropriate that these aspects be studied together.

The design of the reactor building is developed in response to a wide range of requirements, and retention of radioactivity released during an accident is only one dimension.The process to establish building design requirements in general is called Requirements and Functional Analysis. Typically, goals expressed at the highest level must be satisfied by design selections made at that level. These top level design selections then become input requirementsat the next level, and must be analyzed and satisfied by design selection. The process continues downward until all details in the design are specified. Periodically, during the design process, the results must be reviewed from top to bottom and from bottom to top, to ensure that selections made during the process are optimal and consistent with top level requirements.

Through this process, a set of NGNP building functional requirements has been derived.Key drivers for the building design can be grouped in several ways. Spaces within the building are driven primarily by the geometry of the equipment and systems that will be housed within the building. The strength of the building members will be driven primarily by the loads imposed by the equipment and systems, or by other building members. Their thickness is driven primarily by strength or radiation shielding requirements. However, the building is also allocated




functional requirements to protect the equipment and systems within it for both investment protection and to satisfy safety requirements for nuclear plants. The building is allocated the requirement to protect nuclear systems during internal and external hazard events (seismic, tornado missile, aircraft impact, malevolent attack, and others). This study examines the need to allocate a functional requirement to retain radioactive material, and the extent to which it should be embedded to meet this and other requirements.

Based on the information presented in this section, a set of technical and functional requirements for the Reactor Building was developed that incorporates the required and supportive safety functions of the reactor building; the physical security requirements; and other requirements to support non-safety function such as plant operation, maintenance, and access for testing and inspections. These requirements are listed in Table 6.

Table 6 Preliminary Technical and Functional Requirements for NGNP Reactor Building

T&FR Number Requirement

NHSB-1.0 The scope of functions and requirements for the reactor building shall be allocated to the Nuclear Heat Supply Building (NHSB), NHSS HVAC System, and NHSB Pressure Relief System (PRS) as described in the PCDR.

NHSB-2.0 The Nuclear Heat Supply Building (NHSB) shall perform the following functions:

NHSB-2.1 House and provide civil structural support for the reactor and all the SSCs within the scope of the NHSS (building geometry/space)

NHSB-2.1.1 Provide personnel and equipment access for plant construction, maintenance, operation, and inspection of SSCs in the NHSS and for routine and emergency ingress and egress of plant workers

NHSB-2.1.2 Provide radiation shielding for plant workers and the public during normal operation to keep radiation exposures ALARA

NHSB-2.1.3 Limit air flow in neutron fields to keep routine releases of activation products and other releases during normal plant operation from all sources of radioactivity inside the NHSB ALARA

NHSB-2.1.4 Maintain internal environmental conditions (temperature, humidity, air refresh) for NHSS SSCs and operators..

NHSB-2.1.5 Support zoning requirements for HVAC, radiation, fire protection, flood protection, and physical security protection





NHSB-2.1.6 The NGNP shall be designed to physically protect the safety related SSCs in the NHSS from hazards associated with the Hydrogen Production System, Power Conversion System and BOP facilities.

NHSB-2.2 Resist all structural loads as required to support safety functions allocated to the NHSB

NHSB-2.2.1 Provide structural support for the reactor pressure vessel and its internals

NHSB-2.2.2 Provide structural support for the PHTS Helium Pressure Boundary (HPB) and the part of the SHTS HPB inside the NHSB

NHSB-2.2.3 Provide structural support and protection for the NSSS SSCs that contain sources of radioactive material outside the reactor vessel including the FHSS, HSS, and systems containing radioactive waste

NHSB-2.2.4 Protect the NHSB, PHTS HPB and other NHSS SSCs against loads imposed by faults including pipe breaks, chemical releases, fires, seismic failures, and explosions located in the HPS Acid Decomposer Building, SG Building, and other adjacent buildings associated with the HPS, PCS, and balance of plant. Protect the PHTS HPB from loads imposed by SHTS piping resulting from faults including structural failures in the HPS Acid Decomposer building and SG Building

NHSB-2.2.5 Provide structural support for all NHSS SSCs that provide a required safety function for all sources of radioactive material within the NHSB

NHSB-2.2.5.1 Provide structural support for SSCs required for confinement of radioactive material

NHSB-2.2.5.2 Provide structural support for SSCs required to maintain core and reactor pressure vessel geometry

NHSB-2.2.5.3 Provide structural support for SSCs required to control core heat removal including the core, reactor vessel, reactor cavity and RCCS

NHSB-2.2.5.4 Provide structural support for SSCs required for control of heat generation including the reactivity control rods

NHSB-2.2.5.5 Provide structural support for SSCs required for control of chemical attack

NHSB-2.2.6 Provide structural support for all NHSS SSCs that provide a supportive safety function for all identified LBEs as necessary to meet the TLRC (supportive safety function)





NHSB-2.2.7 Provide confinement of radioactive material during normal operation and during LBEs (AOOs, DBEs, and BDBEs) as necessary to meet the TLRC. This function is also supported by the NHSB PRS and NHSS HVAC (supportive safety function) (A design goal of a source term reduction factor of 10 is set in Section A3 for I-131 and Cs-137for releases from the PHTS during design basis events involving DLOFC)

NHSB-2.2.8 Control building leakage and limit air ingress to the core following a large breach or breaches in the PHTS HPB, FHSS, and HSS for all identified LBEs as necessary to meet the TLRC (supportive safety function) (quantification of this requirement to be determined)

NHSB-2.3 Protect the SSCs within the NHSS that perform safety functions from all internal and external hazards as identified in the LBEs as necessary to meet the TLRC

NHSB-2.3.1 Protect the SSCs within the NHSS that perform required safety functions from LBEs involving PHTS and SHTS HPB leaks and breaks. This function is also supported by the NHSS Pressure Relief System.

NHSB-2.3.2 Protect the SSCs within the NHSS that perform required safety functions from LBEs involving missiles from rotating machinery and other internal sources

NHSB-2.3.3. Protect the SSCs within the NHSS that perform required safety functions from LBEs involving internal fires

NHSB-2.3.4 Protect the SSCs within the NHSS that perform required safety functions from LBEs involving internal floods

NHSB-2.3.5 Protect the SSCs within the NHSS that perform required safety functions from LBEs involving hydrogen process hazards

NHSB-2.3.6. Protect the SSCs within the NHSS that perform required safety functions from LBEs involving seismic events

NHSB-2.3.7 Protect the SSCs within the NHSS that perform required safety functions from LBEs involving accident aircraft crashes and transportation accidents

NHSB-2.3.8 Protect the SSCs within the NHSS that perform required safety functions from LBEs involving high winds and wind generated missiles

NHSB-2.3.9 Protect the SSCs within the NHSS that perform required safety functions from LBEs involving other internal and external hazards (requirements to be determined)





NHSB-2.3.10 Protect the SSCs within the NHSS that perform supportive safety functions from LBEs as necessary to meet the top level regulatory criteria as necessary to meet the TLRC (requirements to be determined)

NHSB-2.4 Provide physical security of vital areas within the NHSB against acts of sabotage and terrorism

NHSB-3.0 The NHSB Pressure Relief System shall perform the following functions:

NHSB-3.1 The PRS shall be compatible with the NHSB boundary functional requirement.

NHSB-3.2 The PRS shall be designed, and the NHSB compartments shall be sized so that leaks and breaks up to 10 mm equivalent break size on the PHTS HPB or that part of the SHTS HPB inside the NHSB do not open the PRS so that HVAC filtration capability shall be continuously maintained

NHSB-3.3 PRS shall open to prevent overpressure and thermal damage to any safety related SSC, remove the pressure driving force for radionuclide releases from the reactor building, and depending on the actual design that is chosen, may also be required to reclose to enable post-blow-down filtration for the following design basis event conditions.

NHSB-3.3.1 Breaks in the PHTS HPB in each NHSB compartment containing NHSS piping or vessels up to 100 mm equivalent single ended break size

NHSB-3.3.2 Breaks in the SHTS HPB in each NHSB compartment containing NHSS piping or vessels up to 1 meter equivalent single ended break size

NHSB-3.3.3 Breaks in the HSS piping up to 100 mm equivalent single ended break size

NHSB-3.4 The PRS shall open and prevent excessive damage to the NHSB in response to a Beyond Design Basis 1 meter break in the PHTS piping; excessive damage is defined by exceeding the TLRC using realistic assumptions

NHSB-4.0 The NHSS HVAC System shall perform the following functions:

NHSB-4.1 a) Maintain internal environmental conditions for all NHSS SSCs during normal operation,b) survive conditions for all AOO and DBE LBEs, andc) provide a post-event cleanup function. (supportive safety function)





NHSB-4.2 Maintain environmental conditions for SSCs that perform supportivesafety functions during LBEs (supportive safety function)

NHSB-4.3 Perform radionuclide filtration functions as required to meet the TLRC for all LBEs (supportive safety function)

NHSB-4.4 Perform protective functions to isolate the HVAC during DBE and BDBE HPB breaks to prevent damage from high temperatures and pressures in building compartments during depressurization to enable post blow-down filtration (supportive safety function)




A2 DEFINITION OF ALTERNATIVE REACTOR BUILDING DESIGN CONCEPTS

In order to gain insight into alternative strategies for implementing the safety functions assigned to the NGNP reactor building, a set of alternative design concepts were developed and evaluated. The approach to identifying and evaluating these alternative strategies is comprised of the following steps:

1. Identify both required safety functions and supportive safety functions, as discussed in Section A1

2. Identify alternative design strategies to meet each safety function

3. Define a discrete set of alternative reactor building design configurations with appropriate combinations of features

4. Evaluate each alternative configuration against technical and functional performance criteria, including the following

a. Normal Operating Functional Requirements

b. Investment Protection Functional Requirements

c. Safety Functional Requirements

d. Physical Security and Aircraft Crash Functional Requirements

e. Capital and O&M Cost Impact

f. Licensability

5. Identify a basis for recommending one or more alternative configurations for further development in the Conceptual Design Phase.

Within the scope of this study, most of these dimensions are evaluated on the basis of the PDCR development and engineering judgment. For example, the alternative configurations ofthe reactor building are not expected to have any influence on its ability to provide space or physically support the reactor system equipment. Some features that are included to assess radionuclide retention, such as enhanced building leak-tightness, may impact the ability of operators to gain access to spaces inside the building. This type of impact is assessed by engineering judgment supported by radiological release calculations that are presented in the next section.

The PBMR primary coolant is high-temperature, high-pressure helium. In the event of a leak or pipe break, reactor coolant will fill the reactor building spaces which are normally filled with ambient pressure air. Expansion of the helium and heating of the air will both contribute topressurization of the internal building spaces. There are two key building design strategies




available to address the possibility of internal pressurization: to make the building strong enough to resist the pressure buildup, or to vent the building to limit the internal pressure. In either case, the over-arching requirement is that the building should maintain the geometry and functionality of the reactor shutdown and decay heat rejection systems when subjected to pressure transient loads (in combination with other loads as required by design codes).

If a primary design selection is to vent the building, there are several ways to achieve this goal. These strategies include the creation and maintenance of physical openings in the building boundary and between internal compartments (open vented building). There are several variants to the building vent concept, including closed vents that open when pressure is applied (rupture panels), and openings that re-close after pressure equalizes (re-closable dampers). Vent openings can be equipped with filters to reduce the release of entrained solids.

It follows that the reactor building members must be made strong enough to resist the load applied by the internal pressure. In a vented building design, that pressure may be low for the outermost building boundary, but may still be substantial for some internal structural members during the blow-down. In an un-vented building, the sub-compartment pressurization will still exist, but the outermost boundary of the building will also be subjected to a pressure load. This may require that the building members be made much stronger than would be required without the pressure load. In order to achieve a reasonable design, it may be necessary to make the outer boundary of the building from curved walls and slabs, so that they are not required to resist bending moments.

The reactor building must also protect SSCs from hazards that originate from internal sources. These might include missiles from rotating machinery, ejected parts, and fluid jet impingement forces, as well as fire and flooding events. NGNP will have available all the strategies usually applied to these hazards, including internal walls and slabs to create zoning separation, train separation, flood source isolation, and fire load management. Features in this class are not likely to be influential with respect to radionuclide retention, but they could have some influence on air ingress mitigation.

The reactor building must also protect SSCs from external hazards. Hazards in this class include seismic events, extreme weather and weather-generated missiles, aircraft crash, and resistance to malevolent attack. The design features used to make the reactor building resistant to earthquake, tornado missiles, and aircraft crash are usually sufficient to assure that it is a robust structure. It may turn out that the design needed for protection from external hazards is sufficient to resist internal pressurization without significant additional features. Additionalstrength in the reactor building will usually make it more resistant to internal, external, and pressurization loads. However, adding strength also increases the member size. In order to maintain adequate space for normal functions, the overall structure may increase in size and cost.

Unplanned discharge of reactor coolant into the building will result in the release any radio-nuclides present as circulating activity. Depending on the size of the HPB break, a percentage of the radioactive inventory characterized as dust and plateout activity will also be




released. Design strategies available to enhance retention of these radio-nuclides by the reactor building include both active and passive filtration using HVAC systems designed for normal operation or specifically for post-event cleanup. This strategy can be affected by the building design for zoning and contamination control. It is also possible to enhance the deposition and plate-out that will occur on building surfaces by making the vent pathway tortuous and providing rough surfaces. While this strategy would increase the deposition in the building, it would also increase the sub-compartment peak pressure during the release. Some flow will also occur across the building boundary, in parallel with flow out any vent pathway. Making the building boundary more leak-tight will limit the potential for radionuclide release across the building boundary.

The reactor building may also play a role in limiting the possibility of air ingress after abreak in the HPB. Air ingress could only occur after complete depressurization of the primary loop, and would be inherently limited if the HPB were broken only in one place or at a low point on the system. The extent to which air ingress is a concern is addressed in subsequent sections.The two distinct strategies available to address and limit ingress of air into the core following a pressure boundary failure are introduction of inert gas into the HPB, or introduction of inert gas into the reactor building. If it is deemed necessary to inject inert gas into the HPB, the logical strategy would be to introduce nitrogen using the connections that already exist at the fuel discharge pipes. These connections are normally used to provide cool helium to discharged fuel, but could readily be designed to be connected to a nitrogen source. If it is deemed necessary to inert the reactor building, a system could be provided to release nitrogen or carbon dioxide into the building. The sizing of a building inerting system would be affected by the use of re-closablevent path dampers and by the leak-tightness of the reactor building boundary. Inerting of the building would introduce some safety issues for plant workers.

To assess the need for radionuclide retention by the reactor building, a set of alternative designs have been developed as summarized in Table 7. The strategies to be examined include the following:

• Open vent path, unfiltered

• Filtered, with a controlled leak patho Designed to filter only the delayed source term

o Designed to filter both the prompt and delayed source term

o Participation in the blow-down over a range of sub-compartment volumes

o Variation in building envelope leak tightness

o Addition of expansion volume

• Pressure retaining building with low leak rateo Addition of expansion volume




These strategic design alternatives are driven by specific reactor building functional requirements. The requirement to maintain reactor building geometry leads to a requirement to resist loads imposed during pressure transients. The strength required for the reactor building, and therefore its cost, will be impacted by the vent path and its configuration, or the selection of a pressure retaining building. Rupture panels, re-closable dampers, and filter systems will also play a part in determining the design basis pressure loading. For any given configuration, additional expansion volume will reduce the pressure transient, but may significantly affect the cost.

The strength required to resist pressure transients may not change the design of the building, because other loads, such as seismic, may be controlling. Embedment of the reactor building could reduce the seismic loading, and any building spaces below grade are expected to have an external soil and groundwater pressure that would oppose pressure transient loadings.Seismic and pressure transient loads are not usually combined, because the events are not postulated to occur simultaneously.

Table 7 Alternative Reactor Building Strategies for Performing Safety Functions


No.Design Description

VentedAreaLeakRateVol % /day


Postblow-


Blow-down

release

Delayedfuel

release

1a. Unfiltered and vented

50-100 Open vent No None Passive

1b Unfiltered and vented with rupture

panels

50-100 Internal + External rupture panels

No None Passive

2 Partially filtered and vented with rupture

panels


Yes None ActiveHVAC

3a Filtered and vented with rupture panels



3b Filtered and vented with rupture panels

and expansion volume

25-50 Internal + External rupture panels + expansionvolume






No.Design Description

VentedAreaLeakRateVol % /day


Postblow-


Blow-down

release

Delayedfuel

release

4a Pressure retaining with internal rupture

panels

0.1-1 Internal rupture panels

N/A Passive Passive

4b Pressure retaining with internal rupture

panels and expansion volume

0.1-1 Internal rupture panels + expansionvolume

N/A Passive Passive

The reactor building is also required to protect other safety related components located within the building from events that include pressure transients, internal hazards like fire, and external hazards like seismic events and aircraft crashes. These requirements will generally cause the building to be very robust, and to be divided into rooms and levels by thick walls and slabs. Because the building must be designed to respond to a wide range of required functions, it also has attributes that frequently contribute to its safety performance in ways that were not specifically required. For example, selecting a tortuous vent path to limit the peak pressure loading also provides flow path turns, surfaces, and quiescent regions that will enhance radionuclide retention, even without any requirement. These functions are designated as supportive functional requirements.

There are also a number of design strategies available to enhance or capitalize on supportive safety features. Normal HVAC systems, not required to perform any safety function, can contribute to dose reduction for events that are analyzed realistically, if they are appropriately designed and are protected from event-related damage. The strategies available to enhance HVAC systems include the installation of blast dampers to prevent damage from rapid pressurization, isolation dampers, and physical separation of HVAC trains.

Other features that provide support to achieving safety-related goals, but are not necessarily safety related include systems and design features that can mitigate against radionuclide transport after an accident pressure transport. During normal operation, building features that enhance radiation and contamination control include the design of the HVAC systems, separation of components, and configuration of radiation and contamination control zones. During a blow-down event, passive filtration through engineered filter system or through the building boundary will reduce radionuclide transport. The leak tightness of the building boundary is important to the amount of retention that can be achieved. Following a pressure release, when air and coolant are moving very slowly through the building and across the boundary, the building will also act to impede radionuclide release to the environment. The primary mechanisms for radionuclide reduction in a post-blow-down phase are passive filtration




and deposition on building surfaces, and active filtration by HVAC systems. The design of the building vent pathway and the building leak rate are parameters that can have significant influence on the radiological performance of the building during this phase.

The building also offers some important passive mitigation to air ingress following a HPB break event. The building limits the amount of air available for ingress. It also provides a space that would permit the establishment of an inert gas envelope surrounding the HPB. The building leak rate could be an important parameter affecting this capability. If air ingress is established as a requirement, it will also be possible to introduce inert gas to the HPB via the fuel discharge connections.

The functional requirements for the reactor building apply to all of the alternatives develop for this study. The alternatives have many features in common, including the following:

• Robust structural design to protect against loads from internal and external hazards including seismic events, hydrogen process hazards and physical security against malevolent acts as described in selected licensing basis events.

• HVAC filtration system provided to maintain environmental conditions and provide radionuclide retention ALARA during normal operation and for anticipated operational occurrences (AOOs).

• Blast panels or isolation devices provided to isolate/ protect normal operation HVAC from HPB breaks.

• Physical isolation/separation provided for fuel storage and safety related SSCs from blow-down vent path following HPB breaks.

• Each alternative configuration considered can be combined with all evaluated embedment alternatives.

• Design of the reactor and HPB pressure boundary and passive safety characteristics will limit the potential for air ingress; capability to support ad hoc emergency actions to inert the reactor core cavity via injection of helium or nitrogen is inherent in the PBMR fuel handling system design.

• All designs considered limit the supply of outside air by controlling building leak rate but leak rate varies among the considered alternatives.

• All alternatives have the capability to inject Helium or Nitrogen gas into the core via pipes connected to the Core Unloading Device in the event that this action is deemed necessary in the next phase of the NGNP design. The need for such action has not yet been identified based on current understanding of air ingress phenomena and is not considered further in this study.




A2.1 ALTERNATIVE 1a UNFILTERED AND VENTED

The reactor unit is arranged within the NGNP reactor building such that major components and component groups are placed within interior vaults or sub-compartments and interconnected with high-integrity reactor piping (Refer to Figure 5). Internal vent pathways have been established such that a break in the HPB which occurs in the reactor cavity will expand through a designed opening to the IHX cavity. If a building vent is provided, gas will then flow from the IHX cavity to a building vent vault (shown as the pressure relief pathway), and to the vent itself which connects to the environment. If a break occurs in the core conditioning system (CCS) vault (not shown in Figure 1), it will flow first to the reactor cavity, and then follow the same path as a break in the reactor cavity, to the IHX cavity, building relief pathway, and environment. If a break occurs in the spaces occupied by the fuel handling support systems (FHSS) or helium support systems (HSS) it will be able to flow directly to the building relief pathway. Any primary or secondary helium release in the IHX vault will be able to expand in the IHX cavity and then to the pressure relief pathway and to the environment. Other rooms and spaces within the reactor building will also have some participation in the blow-downpressure transient, because gasses will expand into the rest of the building through penetrations, doors, and other imperfect seals.

Alternate 1a is shown in schematic form in Figure 6. This design alternate assumes that open vent paths interconnect the reactor cavity to the IHX cavity. Open vent paths also connect the IHX cavity and the other rooms containing HPB components to the pressure relief pathway.The spaces that contain the HPB equipment (outlined in black in Figure 1) are designated as the pressurized zone, and designed to have a leak rate of 50-100 vol%/d. This means that the pressurized zone would leak at a rate such that 50-100% of its volume would leak out in one day if it was maintained at its design pressure. Its design pressure will be estimated by performing pressure transient calculations. A carefully designed and constructed, industrial-grade building, with reasonable care taken in the design of doors, windows, and penetrations, should be able to achieve such a leak rate. Gasses may leak from the pressurized zone into the other spaces in the reactor building, which are surrounded by the reactor building envelope (shown in magenta in Figure 6). The reactor building envelope is expected to also have a leak rate on the order of 100 vol%/d. The pressure relief pathway is assumed to be isolated from normal building HVAC systems so that normally required zoning and contamination control can be managed. This alternative is developed for purposes of this study only. It would be difficult to create a design with both open vent paths and isolated HVAC that would function properly. In alternate 1a, it is assumed that reasonable care is taken to create smooth transitions between volumes, avoid high flow loss entrance and exit effects, and minimize the pressure drop along the vent path. In Alternate 1a, there is no feature provided to close the vent path after the blow-down. No filters are provided in this alternative. Radioactivity reduction mechanisms include plate-out,condensation, settling, and decay of radio-nuclides. Following a release event, normal HVAC would be used to provide clean-up, if it is available, but HVAC has no safety function.




Reactor Cavity

IHXCavity

All Other Compartments with PHTS HPB SSCs including HSS and FHSS

Pressure ReliefPathway

PrimaryHTS

SecondaryHTS

Figure 6 Reactor Building Alternative 1a – Unfiltered and Vented

The advantages for alternative 1a include the following:

• Minimizes pressure transient

• Inherently reliable as it is completely passive

The disadvantages for alternative 1a include the following:

• More difficult to control release of air activation products during normal operation

• No engineered features to retain accident related fission products or limit air ingress to building following blow-down

• May require very large HVAC equipment to maintain zoning and zone pressure gradients




A2.2 ALTERNATIVE 1b VENTED AND UNFILTERED WITH RUPTUREPANELS

Alternative 1b is shown in Figure 7 below.

Reactor Cavity

IHXCavity



PrimaryHTS

SecondaryHTS

Rupture Panel - normally closed - opens with internal pressure - does not re-close

Figure 7 Reactor Building Alternative 1b – Unfiltered and Vented with RupturePanels

Alternative 1b is similar to 1a, except that rupture panels have been added to separate internal spaces from each other, and to separate the pressure relief pathway from the environment. The various leak rates are still assumed to be the same as Alternative 1a. Thesepanels are expected to consist of a thin metal or plastic membrane supported on one side. If pressurized from the supported side, the panel is designed to deform and rupture at a low, predictable pressure. There may be devices to aid the rupture process (knife points) that the membrane strikes when it distorts. If pressurized in the opposite direction, against the supports, the membrane will fail at a higher pressure. The introduction of rupture panels will increase the peak pressure transient for all events.

The advantages of Alternative 1b over alternative 1a include the following:

• Easier to control release of air activation products during normal operation

• Easier for HVAC to control pressure zones




The disadvantages of Alternative 1b include the following:

• Increases pressure transient and design pressure

• Rupture panel design will have lower reliability than open vent

• No engineered features to retain accident related fission products

A2.3 ALTERNATIVE 2 PARTIALLY FILTERED AND VENTED WITH RUPTURE PANELS

Alternative 2 is shown in Figure 8, below.

Reactor Cavity

IHXCavity



PrimaryHTS

SecondaryHTS


Reclosable Damper

Filter

Figure 8 Reactor Building Alternative 2 – Partially Filtered And Vented With Rupture Panels

Alternative 2 is based on Alternative 1b, but is provided with two additional features. A re-closable damper and a filter are added in parallel at the point where the pressure relief pathway connects with the environment. In addition, the leak rate for the pressurized part of the building (outlined in black) is reduced to 25-50 vol%/d at 1.1Bar.




The re-closable damper is a device which will open with a small internal pressure, and when the inside and outside pressure have equalized, will re-close. It may be pressure actuated to open and gravity actuated to close, or other mechanism may be used. The filter, preceded by a rupture panel, is provided to reduce the concentration of particulate fission products in any gas stream that flows out of the building after pressure has equalized. The rupture panel preceding the filter is designed to fail open during the initial blow-down, which also opens the damper.The vent path becomes a feature that discriminates between the initial blow-down, which might include prompt source terms, and subsequent releases, which might include delayed source terms.

To achieve a lower leak rate for the pressurized spaces, it will be necessary to upgrade a number of components. Doors and penetrations will probably need to be nuclear grade and to use soft and inflatable seals. The building boundary will need to be sealed at element joints by gaskets, caulking, or welding. These features increase in cost and complexity with increasing design pressure.

The advantages of Alternative 2 include the following:

• Rupture panel on vent paths make it easier to control release of air activation productsduring normal operation, and for HVAC to control pressure zones

• Re-closable damper minimizes pressure transient by opening vent during large HPB break and increases effectiveness of filter

• Filter on vent path mitigates the delayed fuel release source term

The disadvantages of Alternative 2 include the following:

• Rupture panels increase the peak pressure transient

• Damper and filter add to capital cost




A2.4 ALTERNATIVE 3a FILTERED AND VENTED WITH RUPTUREPANELS

Alternative 3a is shown in Figure 9, below:

Reactor Cavity

IHXCavity



PrimaryHTS

SecondaryHTS


Filter

Figure 9 Reactor Building Alternative 3a - Filtered and Vented With RupturePanels

Alternative 3a is similar to Alternative 2, except that it does not have a re-closabledamper on the exit. In this alternative, any reactor building pressurization event must be relieved through the filter or by leaking through the building boundary. This design approach imposes potentially severe design constraints on the filter, and will increase the design pressure. It may be necessary to design the filter for high temperature as well as high flow. Increased design pressure will make the achievement of the leak rate more costly.

The advantages of Alternative 3a include the following:

• Internal rupture panels make it easier to control release of air activation products during normal operation, and control HVAC zones

• Filter reduces both the prompt release source term and the delayed fuel release source term




The disadvantages of Alternative 3a include the following:

• Increased peak design pressure

• Increased cost due to more difficult filter design and higher design pressure.

A2.5 ALTERNATIVE 3b FILTERED AND VENTED WITH RUPTUREPANELS AND EXPANDED VOLUME

Alternative 3b is shown in Figure 10, below.

Reactor Cavity

IHXCavity



PrimaryHTS

SecondaryHTS


Filter

ExpansionVolume

Figure 10 Reactor Building Alternative 3b Filtered and Vented With RupturePanels and Expanded Volume

Alternative 3b is similar to Alternative 3a, except that additional expansion volume is assumed. Such expansion volume could be achieved by increasing the volume of the IHX vault, or by constructing a separate building and connecting it to the IHX vault by a tunnel or passage.Additional volume is expected to reduce the peak pressure for blow-down events. This wouldmake it easier to develop a design for the exit point filter and would reduce the cost of the features needed to achieve the leak rate. The key disadvantage to Alternate 3b would be additional cost related to the increased building space.




A2.6 ALTERNATIVE 4a PRESSURE RETAINING WITH INTERNAL RUPTURE PANELS

Alternative 4a is shown in Figure 11, below.

Reactor Cavity

IHXCavity


PrimaryHTS

SecondaryHTS


Figure 11 Reactor Building Alternative 4a Pressure Retaining With Internal Rupture Panels

Alternative 4a is based on the previous models, but does not have any engineered vent path. It is developed and analyzed to develop insight into the radiological retention performance that a pressure-retaining structure might have when subjected to a range of postulated NGNP accident events. The imposition of a pressure retaining building concept on the NGNP project would probably require a significant revision to the design approach for the reactor building. For analytical purposes, this alternative assumes that the pressure retaining features could be imposed on the pressurized zone within the reactor building. It is likely that the pressure retaining boundary location would be made larger, and located at the building outer boundary, making the NGNP reactor building conceptually similar to a dry containment normally applied to light water reactor designs. This case will also be analyzed.

The leak rate for the pressurized zone in Alternative 4a has been reduced to ~0.1-1vol%/d. To achieve a leak rate this low, it would be necessary to provide nuclear grade airlock doors with inflatable seals and hard (welded) penetrations. It would also be necessary to add a steel or epoxy liner system.




The advantages of Alternative 4a include the following:

• The elimination of the vent path means all radioactive materials released from the HPB are retained within the reactor building

• The low leak rate reduces transport of radio-nuclides to the environment via building leakage.

The disadvantages of Alternative 4a include the following:

• Increased design pressure, with increased costs

• Very low allowable leak rate, with increased costs

• Higher design pressure may require curved walls and slabs

• Higher design pressure may require engineered penetrations, and may require alternative features to maintain passive heat rejection

• Increased inspection and testing required

• Pressure retaining boundary failure modes may have significant consequences (however they are low probability events)

A2.7 ALTERNATIVE 4b PRESSURE RETAINING WITH INTERNAL RUPTURE PANELS AND EXPANDED VOLUME

Alternative 4b is shown in Figure 12, below.

Alternative 4b is based on Alternative 4a, but includes an additional expansion volume.Such additional volume would reduce the design pressure. Lowering the design pressure should reduce the cost, but savings may be overcome by the cost of additional space.

Alternative designs from 1a to 4b have been identified so that the pressure transient response and radionuclide retention functions can be judged along an increasing dimension of building complexity, within the time and budget constraints of the study. These aspects are addressed in Section A3.




PrimaryHTS

SecondaryHTS


ExpansionVolume

Reactor Cavity

IHXCavity


Figure 12 Reactor Building Alternative 4b Pressure Retaining With InternalRupture Panels and Expanded Volume




A3 EVALUATION OF REACTOR BUILDING PRESSURERESPONSE AND RADIOLOGICAL RETENTION

CAPABILITY

A3.1 CHARACTERIZATION OF FISSION PRODUCT TRANSPORT

A depressurized loss of forced circulation (DLOFC) transient consists of three phases:

1. Depressurization Phase2. Heat-up (Expansion )Phase 3. Cool-down (Contraction) Phase

The depressurization phase occurs when the helium coolant within the HPB blows down into the RB and depressurizes the PHTS. During this blow-down the circulating activity that is suspended in the helium atmosphere during normal operation is released to the RB. In addition, some of the activity deposited on the PHTS surfaces in the form of settled dust or plateout activity can be re-suspended and also released to the RB. This re-suspension release increases with increasing break size because a faster blow-down will exert larger liftoff forces on the deposited particles.

The second phase is the heat-up phase where the core fuel heats up after the blow-downis complete and the PHTS is depressurized. During the heat-up phase, a portion of the core reaches elevated temperatures where additional fission products are released primarily from already damaged fuel particles. These radio-nuclides are released in the hot core volume. Due to the temperature increase the helium in the hot volume expands into the cold helium volume by which it is surrounded. By this mechanism some of the radio-nuclides released into the hot volume migrate into the cold volume. The expansion of the hot helium volume into the cold helium volume pushes some of the cold helium volume through the cold break into the RB. The released activity in the RB can settle on the RB surfaces, remain suspended in the RB atmosphere, be trapped on the filters in the RB exhaust or leak out of the RB bypassing the filters.

Other potential radionuclide transport mechanisms between the hot volume, the cold volume and the RB are (1) Buoyancy driven natural convection, and (2) Diffusion. Both of these processes are assumed in this analysis to be insignificant. The helium transport path is very complex, as shown in Figure 13. Helium enters through the cold inlet pipe into the top of the reactor vessel. From there it flows down between the vessel wall and the core barrel to the lower plenum where it turns around and flows up through the side reflector channels. At the top it turns again 180º to flow down through the core to the lower core plenum and out through the hot core outlet pipe to the heat exchangers. The hot core outlet pipe is a concentric double pipe with




the hot helium flowing on the inside and cold helium on the outside. Because of this double pipe arrangement, there are no hot helium pipe breaks in the DBE spectrum.

Figure 13 Helium Flow Path Through NGNP Reactor Pressure Vessel

The helium circulator has a check valve at the outlet that closes on loss of forced circulation and the pipe break for this analysis is assumed to be in the cold pipe between the circulator and the vessel. In order for a natural convection flow to develop, the hot helium, which wants to flow upward because of buoyancy, would have to reverse flow up through the core, down through the reflector and up again along the vessel, and out through the core inlet pipe to the break and out of the break. Because of the closed circulator check valve, the RB atmosphere would have to be drawn in through the same break and flow counter-current to the hot helium through the same flow path to the core in a streamline flow without mixing with the hot helium. This counter-current streamline flow is very unlikely to develop because there is no significant elevation difference between the core and the break location to provide a net




buoyancy force to drive the flow. Even if there was a buoyancy driving force, mixing of the out-flowing hot helium and the in-flowing RB atmosphere would destroy the flow pattern and keep it from developing. Thirdly, the density of helium is 1/7th of the density of air at the same temperature which does not favor the development of buoyancy driven flow. Therefore,buoyancy driven flow has been assumed to have a negligible impact on the radionuclide transport to the RB.

The same complicated flow path also would make diffusion an ineffective means to transport radio-nuclides from the core to the RB and it also has been assumed to be negligible in this analysis. Double pipe breaks of the hot core outlet pipe and the cold shroud pipe and cold pipe breaks with a stuck open circulator exit check valve are BDBEs . For the PRA where BDBEs will be considered the possibility of a buoyancy driven flow may need to be revisited because in the double pipe break the flow pattern would be less complex and in the stuck open check valve case it would not need to be a counter-current flow.

The third phase is the contraction phase where the helium in the PHTS contracts because the fuel is cooling down and the helium temperature in the PHTS is decreasing. During this phase it is possible that air is drawn into the PHTS through the break, and if air comes in contact with hot graphite the graphite can oxidize to form carbon monoxide.

A3.2 TECHNICAL APPROACH

In order to assess the radionuclide retention and pressure capacity of the various Reactor Building (RB) alternatives for a DLOFC, a simplified 3 volume model representing the hot and cold parts of the PHTS and the RB vent volume (RBVV) was developed. The RBVV includes the free volume of the area that would be exposed to temperature and pressure loads following a depressurization event in the PHTS HPB. The model is an integrated model, capable of calculating the pressures and temperatures in each volume during the PHTS blow-down, heat-upand contraction phases of the transient, radionuclide transport between the volumes, radionuclide release from the RBVV and dose estimates at the site 425 m Exclusion Area Boundary (EAB).The heat-up and cool-down of the fuel and the time-dependent radionuclide release during the DLOFC are inputs to the model.

The PHTS is modeled with two volumes, one hot volume to account for the hotter portion of the PHTS, about 25%, and one cold volume to account for the cooler portion of the PHTS, about 75%, during steady state operation. The HPB breaks in this assessment represent breaks in the core inlet pipe (CIP) and are therefore simulated from the cold volume. The two volume model for the PHTS was adopted so that the effect of the time dependent transport of radio-nuclides released during the heat-up in the hot volume to the cold volume could be adequately reflected.

The third volume in the model is used to model the RBVV. The RBVV model is capable of simulating all of the proposed RB alternatives. Radio-nuclides are released from the RBVV




directly to the environment. Deposition and holdup of the radio-nuclides that leak from the RBVV to the remaining RB volume are neglected.

The total air volume of the RB is approximately 100,000 m3 per the NGNP Nuclear Heat Supply Building (NHSB) drawings, Refs. [18]-[21]. The RBVV volume is input to the model as a percentage of the total RB volume. Volumes for the RBVV of 10 to 20 percent of the total RB volume are typical and up to 100 percent of the total RB volume for cases with an added expansion volume. The volume of the 10 percent RBVV case is approximately equal to the volume of the IHX compartment and PRS relief shaft in the NGNP PCDR design.

Each volume in the model has a constant volume throughout the transient. Gas and radio-nuclides transferred between volumes are assumed to mix uniformly and instantly in the volume. The temperature of the hot PHTS volume is based on the heat-up of the fuel and any mixing of the hot and cold volumes during the transient. The temperatures of the cold PHTS and the RBVV are based solely on the mixing of the gases during the transient. Additional heat-upand/or cool-down of the cold volume and the RBVV due to conduction or convection effects are not accounted for in the model.

Insights from previous HTGR studies indicate that I-131 and Cs-137 are key radio-nuclides in determining off-site doses. The model calculates the transport of these two radio-nuclides from the PHTS to the RBVV and then to the environment. The total effective dose equivalent (TEDE) and the thyroid committed effective dose equivalent (CEDE) from inhalation are calculated for each isotope at the 425 m site boundary assuming a ground level release with no deposition accounted for in the environment. The total doses from including all radio-nuclides are then estimated by scaling the I-131 and Cs-137 doses.

The model accounts for the lift-off of plated-out activity and the re-suspension of dust activity in the PHTS during the initial blow-down but neglects the plate-out and resettling of nuclides in the PHTS following the blow-down. Settling in the RBVV is modeled, as is radioactive decay of the I-131 activity in all volumes and the environment. During the heat-up,radionuclides released from the fuel are added to the hot volume and transfer from the PHTS to the RBVV is based on thermal expansion. If the blow-down is complete, once the thermal contraction phase of the PHTS initiates, the release from the PHTS ceases as air ingress from the RBVV begins.

The model also addresses carbon monoxide (CO) formation as the result of air ingress during the cool-down phase of the transient. Two mols of CO are formed for each mol of oxygen that is transferred from the RBVV to the PHTS hot volume. The model only calculates the CO produced as a result of the air ingress from the RBVV into the PHTS due to thermal contraction. A separate analysis addresses air ingress for an assumed fuel inlet pipe rupture at the top of the reactor vessel.




The details of the analytical models and the model input are presented below.1. All pressure differential flows are based on the isentropic choked and un-choked mass

flow rate equations in the MHTGR PSID, Ref. [13]. The discharge coefficient is assumed to be 1.0 in all cases. The flow area for the PHTS leak is calculated by using the leak size as the diameter for a break with a circular area. The effective flow area for the leakage flow from the RBVV is calculated using the leakage rate per day and the corresponding pressure differential for the leakage rate. The effective flow areas for the re-closeable damper and filter damper are input.

2. At equal pressures, the mass flow is based on the ideal gas thermal expansion/contraction of the gas mixture.

3. Lift-off of plated out activity is based on the model in the MHTGR PSID, Ref. [13].

a. 0.2% of deposited Cs-137 and I-131 lift-off for all shear force ratios less than 1.0

b. 15% of deposited Cs-137 lift-off for all shear force ratios greater than 1.0 and less than 30.0

c. 25% of deposited I-131 lift-off for all shear force ratios greater than 1.0 and less than 30.0

d. Shear force ratios greater than 30.0 do not occur in the analyses.

4. The dust re-suspension model based on existing PBMR analyses [14].

a. 0.02% of the Cs-137 dust is re-suspended for all shear force ratios less than 1.0

b. 11% of the Cs-137 dust is re-suspended for all shear force ratios greater than 1.0 and less than 10.0

c. 34% of the Cs-137 dust is re-suspended for all shear force ratios greater than 10.0 5. Settling of activity in the RBVV is based on the model in IAEA TECDOC-978, Ref. [15].

a. Cs-137 deposition constant = 0.1hr-1

b. I-131 deposition constant = 0.3hr-1

6. The TEDE calculation is based on the definition of TEDE in the NRC Reg. Guide 1.183, Ref. [16]. A simplified calculational methodology for the breathing rate, weather factor, and TEDE compared to that described in Ref [16] is used in the model and described in the items that follow. In the model, the TEDE is calculated based on the cumulative release for the 300 hour period following the initiation of the DLOFC.

a. Dose conversion factors are based on EPA Federal Guidance Reports 11 and 12, Refs [22] and [23].

b. A constant breathing rate of 2.3E-4 m3/s based on the >24 hour breathing rate in NRC Reg. Guide 1.183, Ref. [16].

c. A constant weather ? /Q factor of 2.3E-5s/m3 based on 10% of the ground release ? /Q at 425 m for the period of 24 to 96 hours in NRC Reg. Guide 1.4, Ref. [50].




This time period was selected because significant releases from the fuel do not start until >24 hours.

d. The total doses from all radio-nuclides are scaled based on “GT MHR Preliminary Safety Assessment Report, Table 4.4.3-3, Ref. [12].

i. The total thyroid CEDE from all radio-nuclides is assumed to be a factor of 2 greater than the I-131 thyroid CEDE.

ii. The total TEDE from all radio-nuclides is assumed to be a factor of 2.5 greater than the sum of the Cs-137 TEDE and the I-131 TEDE.

7. The calculation of the puff release from a gross RB failure for the pressure retaining RB alternative is based on the point in time when the activity in the RB results in the greatest dose. The product of the breathing rate and weather factor is assumed to be a factor of 18.5 higher than that used otherwise in the dose calculations to account for a sudden release. The factor of 18.5 is based on the breathing rates and 10% of the 0- to 8-hourweather factors in the NRC Reg. Guide 1.4, Ref. [50].

8. The time step during the calculation is 0.01 s for all times when the PHTS pressure is greater than the atmospheric pressure and 100 s for times when the PHTS pressure is equal to the atmospheric pressure.

9. The RBVV helium/air mixture properties are calculated at each time step.

10. The model inputs for each volume and for the environment are based on existing PBMR calculations for the 500 MWt NGNP design and are given in Table 8.

11. The time dependent average fuel temperature during the DLOFC is input to the model and shown in Figure 14. The source for this data is from existing PBMR calculations for the 500 MWt NGNP design. The average fuel temperature data was only provided for 0to 72 hours. Based on results from existing calculations for a 500 MWt design with a different operating condition, the average fuel temperature is assumed to decrease linearly to about 750 ºC between 72 and 300 hours. The peak average fuel temperature of about 1100°C occurs at about 43 hours. During the transient, 28% of the PHTS hot volume is assumed to follow this temperature profile.

12. The characterization of the initial radiological inventories and the releases from the fuel during DLOFC are discussed in Sections A3.5 and A3.6, respectively.

13. The model input for the re-closeable damper, filter damper and filter is given in Section A3.4 for each alternative.




Table 8 NGNP Initial Conditions for Model

NGNP Plant Parameter Value

Mass of PHTS cold volume, kg 2798.89

Volume of PHTS cold volume, m3 376.43

Temperature of PHTS cold volume, °C 295.19

Mass of PHTS hot volume, kg 472.53

Volume of PHTS hot volume, m3 117.23

Temperature of PHTS hot volume, °C 798.02

Pressure of RBVV, Pa 1.013x105

Temperature of RBVV, °C 20

Pressure of environment, Pa 1.013x105

Temperature of environment, °C 20

Steady state mass flow rate of PHTS, kg/s 160

NGNP Average Fuel Temperature during DLOFC

600

700

800

900

1000

1100

1200

0 50 100 150 200 250 300

Time (hr)

Tem

pera

ture

(C)

Figure 14 NGNP Average Fuel Temperature during DLOFC




A3.3 KEY ASSUMPTIONS

There is no integrated model available that is capable of modeling all of the aspects of the DLOFC including the blow-down, heat-up and contraction of the PHTS, the radionuclide release from the fuel, the transfer of radio-nuclides from the PHTS to the environment and the dose calculations for the NGNP design. Given the schedule and budget constraints for the project, a number of major assumptions were made in order to evaluate the RB alternatives.

1. Only 3 volumes are used to model the PHTS and RB.

2. Use of the DLOFC average fuel temperature to vary the hot volume temperature during the transient and the lack of additional heating or cooling of the PHTS cold volume or RBVV volume during the transient due to heat transport mechanisms other than mixing.

3. The radionuclide release from the fuel during the DLOFC is based on existing calculations for a 500 MWt PBMR design at different steady state operating conditions.The same radionuclide release profile is used for leaks of all sizes. These delayed releases are considered conservative for break sizes less than 100 mm due to the enhanced convection cooling of the core which lowers the peak temperatures and thereby lowers the release from the fuel during the heat-up.

4. The initial radionuclide inventories are taken from the NGNP Contamination Control Report, Ref. [10].

5. Use of simplified liftoff and dust re-suspension models based on the shear force ratio calculated by the model.

6. Mass transfer between the PHTS volume, the RBVV and the environment based on the pressure blow-down and on expansion/contraction during the heat-up and cool-downphase of the DLOFC.

7. No convection or buoyancy driven flow after depressurization.

a. All DBE breaks in the PHTS outside the reactor vessel are breaks from the cold volume.

b. The circulator discharge check valve closes at the end of the blow-down phase at the latest.

c. The flow path for buoyancy driven convection from the core to the break location requires flow reversal and counter-current flow with multiple up-down paths.

d. Flow patterns are complex and do not favor buoyancy driven convection.

8. Simplified dose calculational methodology as outlined in Section A3.1. All releases are treated as ground releases.




A3.4 BOUNDARY CONDITIONS

For the assessment of the RB pressure capacity, a design basis event involving a 1000 mm break (equivalent break size for a double-ended guillotine break (DEGB) of the SHTS HPB) is assumed due to the larger helium inventory of the SHTS relative to the PHTS. The lack of cooling for the SHTS based on the PCDR design introduces large uncertainties in the reliability of the hot gas pipe liner and insulation whose failure would directly lead to a major rupture of the SHTS HPB. Hence this break was classified as a design basis event, whereasbreaks larger than 100 mm in the PHTS HPB are classified as beyond design basis events.

For the assessment of the RB radionuclide retention, the selected DBEs are limited to Depressurized Loss of Forced Cooling (DLOFC) cases for a range of break sizes in the core inlet pipe (CIP). Equivalent break sizes of 2, 3, 10, 100, 230, and 1000 mm are analyzed. The DEGB of the 710 mm CIP has an equivalent break size of 1000 mm.

Offsite doses for the DLOFC events at the 425 m site boundary are compared to the Top Level Regulatory Criteria of the Nuclear Regulatory Commission (NRC) and the Environmental Protection Agency (EPA) for assessing and regulating nuclear power plants in the US. These requirements are shown in Table 9. The TEDE is the sum of the CEDE from inhalation and the effective dose equivalent (EDE) from external exposure. The thyroid CEDE is the committed effective dose equivalent to the thyroid from inhalation. The TEDE and thyroid CEDE limits are based on the contributions from all radio-nuclides.

Table 9 Top Level Regulatory Criteria used to Evaluate Site Boundary Doses

Regulation and Reference Application Offsite Dose Limits

NRC 10CFR50.34 and NRC 10CFR50.67, Ref. [8]

Offsite dose limits during design basis events (DBEs) TEDE < 25 rem/event

EPA Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, Ref. [9]

Offsite dose limits during DBEs and beyond design basis events (BDBEs) at which sheltering is considered

TEDE < 1 rem/event

Thyroid CEDE < 5 rem/event

A3.5 CHARACTERIZATION OF INITIAL RADIONUCLIDE INVENTORIES

The I-131 and Cs-137 source term inside the PHTS that is available for release at the beginning of the DLOFC is shown in Table 10 and is based on the NGNP Contamination Control Report, Ref. [10]. The plate-out and dust activities in Ref. [10] are based on the assumption of 60 years of continuous full power operation with an availability of 95 percent. No credit was




taken for changes in the system conditions related to power operation and maintenance activities.For the model input, the Cs-137 plate-out and dust activities in each volume were reduced by 40percent to be consistent with the total possible lifetime activity based on the Cs-137 steady state release rate given in Ref [10].

Table 10 Initial Release Source Terms for I-131 and Cs-137 Radio-nuclides

Source Activity (Ci)I-131 PHTS cold volume circulating 0.0I-131 PHTS hot volume circulating 0.0I-131 PHTS cold volume plateout 0.13I-131 PHTS hot volume plateout 3.48Cs-137 PHTS cold volume circulating 0.0Cs-137 PHTS hot volume circulating 0.0Cs-137 PHTS cold volume plateout 6.18Cs-137 PHTS hot volume plateout 191Cs-137 PHTS cold volume dust 4.67Cs-137 PHTS hot volume dust 259

A3.6 MATRIX OF RADIOLOGICAL EVALUATION CASES

Using the three volume model described above, a matrix of cases was evaluated to determine the radiological consequences and the pressure response of the reactor building alternatives. For the DBEs a range of break sizes ranging form 2 mm to 100 mm equivalent diameter in the core inlet pipe were analyzed with reactor building vent volumes ranging from 10 to 100 percent of the nominal reactor building volume of 100,000 m3. In all cases a total loss of forced circulation (LOFC) was assumed concurrent with the break in the Primary Heat Transport System (PHTS).

For the BDBE evaluation of margins and limiting reactor building pressures, two leak sizes were considered: a 230 mm equivalent break size, which corresponds to partial break in the 710 mm Core Inlet Pipe (CIP), and a 1000 mm equivalent break corresponding to a double ended guillotine break of the CIP. The same range of reactor building vent volumes (RBVV) from 10 to 100 percent was considered.

The reactor building alternatives considered were (1) no reactor building and (2) the reactor building alternatives 1a, 1b, 2, 3a, 3b, 4a and 4b described in Section A2.

In addition two special cases were considered: (1) Alternatives 4a (puff) and 4b (puff) are the same as Alternatives 4a and 4b with an assumed gross reactor building failure simulated by a puff release of the activity in the blow-down release from the reactor building at the worst point




in time. This alternative was considered because in Alternative 4 the entire radionuclideinventory is retained in the RB in a pressurized state which would be available for a RBdepressurization release in the event of a containment failure due to external causes (seismic,etc.) after the radionuclide release from the core has occurred. (2) A 1000 mm equivalentdiameter SHTS break was considered as the limiting RB pressure response because the SHTShelium inventory is somewhat larger than the PHTS.

This case matrix analysis not only represents an evaluation of the previously described design alternatives, but the results can also be interpreted to represent a spectrum of scenarios in which the various RB design features (rupture panels, filters, pressure retaining features) are successful and unsuccessful in performing their respective safety functions such as would be developed in a full PRA. For example, the Alternative 1a results would also be applicable to Alternative 2 or 3a with failure of the filters and relief shaft isolation dampers.

For the analytical model the RB Alternative 2 shown in Figure 15 is used because it allows modeling of all the RB alternatives by varying the set-points of the re-closeable dampers and filter dampers. The different RB Alternatives are simulated with the inputs shown in Table11.

Reactor Cavity

IHXCavity



PrimaryHTS

SecondaryHTS


Reclosable Damper

Filter

Figure 15 RB Alternative 2 Used as Basis for the RB Analytical Model




Table 11 Inputs Used for the RB Model to Simulate the Different RB Alternatives

RBAlter-native

LeakRate

[vol%/day)

LeakRate

Pressure[bar-d]

RBVV[m3]

RecloseDamper

openpressure

[bar-a]

RecloseDamper

closepressure

[bar-a]

RecloseDamper

Flowarea[m2]

FilterDamper

Openpressure

[bar-a]

Filterdamper

Flowarea[m2]

FilterDecon.Factor

Cs [%]

FilterDecon.FactorI [%]

1a 100 0.1 10,000

20,000

N/A N/A N/A N/A N/A N/A N/A

1b 100 0.1 10,000

20,000

1.113 N/A N/A N/A N/A N/A N/A

2 50 0.1 10,000

20,000

1.213 1.113 3 [Note 1] 3 99 95

3a 50 0.1 10,000

20,000

N/A N/A 3 1.213 3 99 95

3b 50 0.1 50,000

100,000

N/A N/A 3 1.213 3 99 95

4a 1 10 10,000

20,000


4b 1 10 50,000

100,000


[Note 1] Filter damper opens when re-closable damper re-closes

A3.7 CHARACTERIZATION OF RELEASES FROM FUEL DURINGDLOFC

The delayed radionuclide release from the fuel during the DLOFC is a function of the temperature during the transient. The calculations for the fuel release assume an atmospheric pressure at the beginning of the transient. This assumption is valid for a rapid depressurizationcases. These delayed releases are considered conservative for break sizes less than 100 mm due to the enhanced convection cooling of the core which lowers the peak temperatures and thereby lowers the release from the fuel during the heat-up transient. As a result, releases from the fuel for breaks up to 100 mm are overstated in this analysis compared with a realistic assessment.However, this conservatism is acceptable for the purposes of this analysis: namely to compare alternative mitigation strategies for the reactor building design.

The temperature distribution profile during the transient is shown in Figure 16 and is based on existing calculations for a 500 MWt PBMR design at different steady state operatingconditions than those for the NGNP [11]. Only a small amount of the fuel reaches temperatures




above 1,700°C during a DLOFC transient, and only for a period of about 40 hours. The corresponding time dependent cumulative I-131 and Cs-137 releases from the fuel during this DLOFC are given in Figure 17 and are input for the model. The maximum I-131 release from the fuel during the transient is about 1,040 Ci and occurs at about 72 hours after which the decay of the source term is more than the quantity released. The maximum Cs-137 release from the fuel is 52 Ci, of which 90 percent is released by 72 hours. A variation from the PBMR DLOFC I-131 source term is that the model source term is used as the cumulative release and then decayed in the model. This assumption underestimates the I-131 maximum release somewhat but is considered reasonable in view of other model uncertainties.

PB Volume Distribution during DLOFC

0

10

20

30

40

50

60

70

80

90

100

0 40 80 120 160 200 240

Time (hr)

% P

B V

olum

e

>300 °C >400 °C >500 °C >600 °C >700 °C >800 °C >900 °C >1000 °C >1100 °C

>1200 °C >1300 °C >1400 °C >1500 °C >1600 °C >1700 °C >1800 °C

Figure 16 Core Fuel Time at Temperature Distribution during DLOFC




0.1

1.0

10.0

100.0

1,000.0

10,000.0

0 50 100 150 200 250 300

Time (hr)

Act

ivity

(Ci)

Cs137

I131

Figure 17 I-131 and Cs-137 Cumulative Activity Release from Fuel during DLOFC

A3.8 EVALUATION OF SITE BOUNDARY DOSES FOR THE NO REACTOR BUILDING CASE

The no RB case is an important case to consider because it sets the radionuclide mitigation required to be provided by any RB alternative. The release of radionuclide release in this case is directly from the HPB to the environment. The analysis of the break size range from 2 mm to 1000 mm showed that for all cases the 3 mm break size is the limiting break size. The reason for this is discussed in the following section.

The results for the no RB case areas are as follows:

TEDE at the site boundary (425 m): 0.4 remThyroid dose at the site boundary (425 m): 10.0 rem

The no RB TEDE represents 1.6 percent of the 10CFR50.34 dose limit and 40 percent ofthe PAG dose limit for DBEs. No mitigation by a RB is needed to meet either TEDE limit.




The Thyroid dose is 2 times the PAG dose limit, and therefore some mitigation by a RB is required to meet the PAG dose limit. Note that meeting the PAG limit at the site boundary is an NGNP user requirement to avoid the need for an emergency planning zone beyond the site boundary and it is not a regulatory requirement.

A3.9 EVALUATION OF SITE BOUNDARY DOSES FOR REACTORBUILDING ALTERNATIVES

As for the no RB case, the 3 mm break size is the dose limiting size for all RB alternatives. The reason for this is illustrated in Figure 18 for RB Alternative 1a. Figure 18 shows the TEDE for DLOFC as a function of break size. This figure is based on a selective, but limited, number of break size cases and is useful in explaining the trends in doses versus leak size; however, it may not represent the exact doses at all leak sizes in between those analyzed.The curve shows a minimum dose at a break size between 10 and 100 mm. At break sizes greater than 100 mm, the shear force ratio during the blow-down increase sufficiently that more of the radio-nuclides deposited in the PHTS are released from the PHTS and the dose increases accordingly. For break sizes smaller than about 10 mm, the PHTS is depressurizing during the time of significant release from the fuel providing a greater transport mechanism for the release from the PHTS to the RB. For the larger leak sizes the PHTS is depressurized before the heat-uprelease and the only mechanism available for release of the radio-nuclides from the PHTS to the RB is thermal expansion. For break sizes in the 3 mm to10 mm range, the smaller the break size, the more the blow-down overlaps the heat-up release and the more radio-nuclides are transported to the RB by the continuing blow-down. At break sizes less than 3 mm, the blow-down time is long enough that the decay of the I-131 while still inside the PHTS HPB becomes a significant factor in reducing the doses. Therefore, the 3 mm break size is the controlling break size for the DBE break size spectrum.

Table 12 shows the dose results for the 3 mm break size case for the RB alternatives with a 10% RB vent volume. The table gives the TEDE and the thyroid dose. The thyroid dose is more limiting when compared to the PAG limit than the TEDE for the 3 mm break size for all alternatives. The margin factor given in the last column is defined as the PAG thyroid dose limit (5,000 mrem) divided by the calculated dose. As shown in the table, the margin factor is significant, exceeding a factor of 10 in all cases.




1.0E-02

1.0E-01

1.0E+00

1.0E+01

1.0E+02

1.0E+03

1.0E+04

1.0E+05

1 10 100 1000

Break Diameter [mm]

Site

Bou

ndar

y D

ose

TED

E [m

rem

]

Direction of increasing Shear Forces during Blowdown to lift-off plateout and dust

Direction of increasing effect of blowdown

driving out the delayed fuel release



From 3mm to 2mm blowdown time is

greater than time of fuel release allowing more decay before release

Figure 18 Dose Behavior vs. Break Size for RB Alternative 1a with 10% RB Vent Volume

Table 12 Dose Results for the RB Alternatives with a 10 % RB Vent Volume

RBAlternative

Leak Size [mm]

TEDE Dose [mrem]

Thyroid Dose [mrem]

MarginFactor

1a 3 16 400 12.5 (Th)1b 3 16 400 12.5 (Th)2 3 1 20 250 (Th)

3a 3 1 20 250 (Th)4a 3 1.4 33 150 (Th)

Table 13 shows the dose results and margin factors again for the limiting 3 mm case for the large RB vent volume alternatives and for Alternatives 4a/b (puff) with the puff release simulating a gross RB failure for the pressure retaining RB alternative. For RB Alternatives 3a/3b, the margin factor increases in proportion to the RB vent volume. Since the maximum




pressure during the transient is the same in these cases and controlled by the filter opening set-point, the amount of radionuclides transferred to the RB is the same for all the 3a/3b cases. The release from the RB decreases as the vent volume increases due to the lower concentration of the radio-nuclides in the larger volume. For Alternatives 4a/4b, the increased RB vent volume results only in a small increase of the margin factor. In Alternatives 4a/4b, there are competing effects as the vent volume increases. The release from the RB decreases as the vent volume increases as in the 3a/3b alternatives, but the amount of radio-nuclides that are transported to the RB is greater as the vent volume increases. The maximum RB pressure is different for all of the 4a/4b cases and decreases as the vent volume increases. Therefore, the largest vent volume case has the largest pressure decrease in the blow-down therefore it has the largest radionuclide transfer from the PHTS to the RB. For Alternative 4a/b (puff), the margin factor is less than one.For this alternative both the blow-down release and the heat-up release are stored within the RB and PHTS at an elevated pressure. When the gross RB failure for the pressure retaining RB alternative occurs and the RB blows down to atmospheric pressure, a significant portion of the stored radio-nuclides are released. This pressure driven release would not occur in the other RB alternatives because the RB is already at atmospheric pressure.

Table 13 Dose Results for the RB Alternatives with a large RB Vent Volumes

RBAlternative

RBVV[m3]

Leak Size [mm]

TEDEDose

[mrem]

ThyroidDose

[mrem]

MarginFactor

3b 50,000 3 0.2 4 12504b 50,000 3 1.3 30 1673b 100,000 3 0.1 2 25004b 100,000 3 1.1 25 200

4a (puff) 10,000 3[note 1] 723 18,000 0.284b (puff) 100,000 3[note 1] 943 23,000 0.22[Note 1]: The accumulation of source term in the building prior to the gross buildingfailure and release is from a 3 mm break of the PHTS




Table 14 lists more detailed results for RB Alternative 1a. In addition to the doses the table gives the maximum RB vent volume pressure, the time to depressurize and the time when the air ingress starts.

Table 14 Detailed Results for RB Alternative 1aRB

AlternativeRBVV[m3]

LeakSize[mm]

TEDEDose

[mrem]

TEDE[mrem]

MaxRBVV

Pressure[bar]

Time to Depressurize

to 1atm

Start of Air

Ingress

1a 10,000 1000 7.7 38 4.3 16.2 sec 43 hr1a 10,000 230 6 15 1.8 63 sec 43 hr1a 10,000 100 0.1 0.1 1.125 334 sec 43 hr1a 10,000 10 4.9 0.3 1.013 9.3 hr 43 hr1a 10,000 3 398 16 1.013 102 hr 102 hr1a 10,000 2 326 14 1.013 230 hr 230 hr1a 20,000 1000 6.5 32 2.8 28.6 sec 43 hr1a 20,000 100 0.07 0.08 1.115 334 sec 43 hr1a 20,000 3 206 8.5 102 102 hr 102 hr

Some insights from the Alternative 1a analysis are as follows:

• The initial I-131 inventory for release in the blow-down is small, the major I-131 activity source is the activity release during heat-up.

• The maximum I-131 release from the fuel during the transient occurs at about 72 hours after which the decay of the source term is more than the quantity released. The maximum Cs-137 release occurs at 300 hrs of which 90% is released by 72 hours. 1% of the total release from the fuel is reached at 11 hours for I-131 and at 15 hours for Cs-137.

• For the 10 mm break size the depressurization is complete at 9.3 hours. For break sizes greater than 10 mm the depressurization is complete before a significant release from the fuel occurs.

• For break sizes greater than 10 mm the doses are low because the release from the fuel is stagnant in the hot volume without a driving force to the environment except for expansion until the cool-down starts.

• For break sizes greater than 10 mm the air ingress into the PHTS starts at 43 hours, the time of PHTS cool-down and is not perturbed by the blow-down. For break sizes less than 10 mm the blow-down extends into the cool-down phase and delays the air ingress to 102 hours for a 3 mm leak and to 230 hours for a 2 mm leak.

• For break sizes less than 10 mm the extended blow-down provides a significant convective mechanism to drive radio-nuclides out of the HPB during the time of the heat-up release from the fuel which is the major release phase.




• The major doses result for break sizes less than 10 mm because of the extended blow-down transport of radio-nuclides to the RB vent volume.

• The slow blow-down rates allow major retention of radio-nuclides in the RB vent volume.

• Doubling the size of the RB vent volume reduces the maximum dose by almost a factor of two.

Figure 19 and Figure 20 show the time dependent release of I-131 for Alternative 1a with a 10% RB vent volume for a 3 mm break and a 2 mm break, respectively. These figures show the release from the fuel, the activity retained in the HPB and in the RBVV and the activity released from the RB. Due to the long blow-down duration most of the activity released from the fuel is retained in the RB vent volume due to the lack of a thermo hydraulic driving force for its release.

1.0E-04

1.0E-03

1.0E-02

1.0E-01

1.0E+00

1.0E+01

1.0E+02

1.0E+03

1.0E+04

0 50 100 150

Time from HPB Break (hrs)

I-131

Rel

ease

d (C

urie

s)

Fuel Release

Release from RB

Retained in HPB

Retained in RBVV

PHTS Fully Depressurizesat 102hrs

Figure 19 Time dependent release of I-131 for a 3 mm Break in Alternative 1a with a 10% RB Vent Volume




1.0E-04

1.0E-03

1.0E-02

1.0E-01

1.0E+00

1.0E+01

1.0E+02

1.0E+03

1.0E+04

0 50 100 150 200 250


I-131

Rel

ease

d (C

urie

s)

Fuel Release

Release from RB

Retained in HPB

Retained in RBVV

PHTS Fully Depressurizesat 230hrs

Figure 20 Time dependent release of I-131 for a 2 mm Break in Alternative 1a with a 10% RB Vent Volume

Figure 21 compares the I-131 release for the 2 mm Break and the 3 mm Break from the two previous figures, but on a linear scale. It shows that the I-131 release in the 3 mm break is significantly higher because the extended blow-down time in the 2 mm break allows more of the I-131 to decay.




0

10

20

30

40

50

0 50 100 150 200 250


I-131

Rel

ease

d (C

urie

s)

RB Release - 2mm

RB Release - 3mm

End of PHTS Depressurization and release from RB

Figure 21 Comparison of the I-131 Release for the 2 mm Break and the 3 mm Break for Alternative 1a

Figure 22 compares the TEDE as a function of the break size for the different RB alternatives. All four RB alternatives clearly meet the TEDE limits. In the DBE domain, for breaks less than 100 mm, the Alternatives 2 and 3a are clearly superior to Alternatives 1a and 4a.In the BDBE domain, for breaks greater than 100 mm, the Alternatives 3a and 4a are clearly superior to Alternatives 1a and 2a. Overall, alternative 3a provides the best radiological protection over the entire break size spectrum for the alternatives shown in Fig 22. Not shown is Alternative 3b, which has a larger assumed RBVV and its radiological retention performance is the best of all the alternatives that were analyzed in this study.

Figure 23 compares the thyroid dose for the different RB alternatives with a 10 % RB vent volume for the 3 mm break size. As discussed earlier the no RB case exceeds the thyroid PAG limit by a factor of two. All RB design alternatives are well below the thyroid dose limit except for alternative 4a with the gross RB failure for the pressure retaining RB (Alternative 4c), which exceeds the PAG thyroid dose limit by a factor of four. Design Alternatives 2 and 3a provide the most margin to the PAG thyroid dose limit. The thyroid dose for Alternative 4c, representing gross RB failure for the pressure retaining RB, exceeds that for the no reactor building case because the release from alternative 4c is a prompt release compared to a slow release for the no RB case. The product of the breathing rate and weather factor is a factor of 18.5 greater for the prompt release.




1.0E-03

1.0E-02

1.0E-01

1.0E+00

1.0E+01

1.0E+02

1.0E+03

1.0E+04

1.0E+05

1 10 100 1000

HPB Break Size [mm]

Site

Bou

ndar

y D

ose

TED

E [m

rem

]

Alternative 1a/1bAlternative 2Alternative 3aAlternative 4a



Figure 22 Dose vs. Break Size for the RB Alternatives with 10% RB Vent Volume




10

100

1,000

10,000

100,000

No RB Alternative1a/1b

Alternative 2 Alternative 3a Alternative 2 or3a with failure ofadded features

Alternative 4a Alternative 4awith late CF

Site

Bou

ndar

y Th

yroi

d D

ose

[mre

m]

EPA PAG Limit = 5 rem Thyroid

Figure 23 Thyroid Dose for the alternative Designs with a 10% RB Vent Volume for the 3 mm Break Size

A3.10 SUMMARY OF RESULTS FOR REACTOR BUILDINGPRESSURE RESPONSE

For DBEs the RB vent volume peak pressure is controlled by the 100 mm leak and by the RB vent volume leak rate. Table 15 summarizes the results for the RB peak pressure for the different RB alternatives and RB vent volumes.

For RB alternatives 1a and 1b (leak rate 100 volume %/day at 0.1 bard), and for RB alternatives 3a and 3b (leak rate 50%/d at 0.1 bard) the RB vent volume peak pressure is below 0.125 bard.

For RB alternative 2 (leak rate 50%/d at 0.1 bard) the RB vent volume peak pressure is controlled by the opening pressure of the reclosable damper (0.2 bard).




For RB alternative 4a and 4 b (1%/d at 10 bar), the peak pressure ranges from 1.4 bar to 5.1 bar, decreasing with increasing RB vent volume. For this alternative there is a significant tradeoff between design pressure and volume.

For beyond design basis break sizes, the RB would experience high RB vent volume pressures, up to 4.3 bar for Configurations 1a, 2 and 3a with a 10% RB vent volume in a 1000 mm cold leg HPB break DLOFC. Therefore, with a low RB vent volume design pressure to meet the DBE needs, the RB would be likely to fail on overpressure for the BDBE break sizes and therefore the release might behave more like a No RB Configuration.

Table 15 Design Basis Peak Pressures for RB Design Alternatives

RB Alternative RBVVm3

Max RBVV Pressure[bar]

1a 10,000 1.1251a 20,000 1.1151b 10,000 1.1381b 20,000 1.1312 10,000 1.213[Note 1]

2 20,000 1.213[Note 1]

3a 10,000 1.1253a 20,000 1.1163b 50,000 1.0973b 100,000 1.0794a 10,000 5.14a 20,000 3.14b 50,000 1.94b 100,000 1.4

Notes: 1. Re-closeable damper opening pressure

The more limiting condition for the RBVV design pressure would be the maximum size break of the SHTS pipe of 1000 mm, a DBA. Due to the larger secondary side helium inventory, the peal pressures in the RBVV are higher than for the 1000 mm PHTS breaks. The SHTS has a total helium inventory mass of 3423 kg, which is 5% higher than the PHTS mass.

The peak pressures for the 1000 mm SHTS hot leg break were analyzed for Alternative 1a, but the 100 mm results are the same for all design alternatives, except for design alternative 4. The results for design alternative 1a are (as shown in Table 16):




Table 16 Peak Pressures for the 1000 mm SHTS Hot Leg BreakRBVV

(% of RB Volume)Peak RBVV Pressure

(bar)Time of Peak Pressure(s)

10 5.1 1.920 3.3 2.550 2.0 3.2100 1.5 3.5

These peak pressures are about 25% higher than the corresponding 1000 mm PHTS breaks, and the peak pressure occurs at only a few seconds.

A3.11 AIR INGRESS

Air ingress is of interest because at elevated temperatures graphite can oxidize in thepresence of air (oxygen) and form carbon monoxide, a flammable gas and the reaction is exothermic. The reaction of interest is:

C + ½ O2 => CO + 110.5 kJ/mol-CO

Two potential air ingress paths were considered in this study to determine whether air ingress could influence the RB requirements. (1) air ingress and CO formation during the cool-down/contraction phase of a cold leg HPB break DLOFC, and (2) air ingress through a fuel inlet pipe break at the top of the reactor vessel.

(1) Air ingress and CO formation during the cool-down phase of a cold leg HPB break DLOFC

Two air transport mechanisms were considered in this study: (a) buoyancy driven counter-current flow, and (b) Primary system contraction during the cool-down phase.

(a) Buoyancy Driven Counter-Current Flow

Air can theoretically be drawn into the reactor vessel by a buoyancy driven flow. Hot helium in the core wants to rise and push cold helium out through the break and draw the cold air-helium mixture from the RB into the vessel through the break. The closed circulator discharge check valve forces hot helium and cold air to flow in counter-currentstreamline flow past each other without mixing.

The flow path is very complex. For breaks downstream of the circulator discharge check valve, the helium has to flow from the hot core up to the core inlet plenum, then down through the outer reflector to the cold lower plenum, then up along the core barrel to the inlet plenum at the top, then out through the cold inlet pipe to the break before the checkvalve (see Figure 13). The cold helium-air mixture from the RB must be drawn in through the break and flow counter-current to the hot helium without mixing.




For breaks upstream of check valve, the hot helium has to flow against the buoyancy forces down from the hot core to the hot lower core plenum and out through the hot outlet pipe to the break between the heat exchangers or between the second heat exchanger and the circulator. The cold helium-air mixture must be drawn in through the break and flow counter-current to the hot core without mixing.

Streamlined counter-current flow without mixing in these complex geometries are very unlikely. The buoyancy force in hot helium is 1/7th of same temperature buoyancy in air because of the low He density, making the development of a buoyancy driven flow even more unlikely. This transport mechanism for air was assumed to be negligible in this study.

(b) Primary System Contraction during the Cool-down Phase

Air ingress does not start until 43 hours, and even longer for breaks < 10 mm. The contraction of the helium at atmospheric pressure in the PHTS draws the cold air-heliummixture from the RB through the break into the cold PHTS volume where it mixes with helium in the cold PHTS volume. Contraction of the hot PHTS volume draws the air-helium mixture in the cold volume into hot volume. Oxygen reaching the hot core can react with hot graphite. Calculations with the 3 volume model show that less than 2 mols of CO, a negligible amount, is produced by 300 hours due to contraction effects alone.Since cooling mechanisms other than gas mixing in the volumes is neglected in the model, the CO produced by contraction may be underestimated. Since the CO quantities are so small, it is estimated that the additional cooling will also yield small quantities.

(2) Air Ingress through a Fuel Inlet Pipe Break at Top of Reactor VesselThere are three fuel tubes that rise in a vertical pipe from the core to the top of the reactor vessel (one fuel tube per 120 º core sector). The fuel tubes are 65 mm diameter with ribs to guide the 60 mm diameter fuel pebbles. A double ended guillotine break at the top of the vessel with a concurrent total loss of forced circulation is postulated. The blow-downof first the hot helium in the core and then the cold PHTS volume through the core will tend to cool the core down. The PHTS is depressurized at 1000 seconds. After the blow-down hot helium in the core will rise through the fuel tube and draw the cold air-heliummixture from the RB into the core. This is a reasonably optimal configuration for counter-current buoyancy driven flow to develop with hot gas at the bottom and cold gas at the top connected by a straight vertical tube.

Two equations must be satisfied:

1. Force Balance: Buoyancy force = Friction losses

2. Volume Flow balance: Volumetric flow rate of helium out = air in

The set of equations was solved by a double iteration on the friction factors for helium and for air under the following conservative assumptions:




• At the end of the blow-down it is assumed that the helium temperature in the core and the air-helium mixture in the RB are both at 1000 ºC.

• Assume pure air the in RB. In reality the RB atmosphere will be mostly helium because the air has been displaced in the blow-down. This assumption is conservative and maximizes the buoyancy.

• Assume 8 ribs of 2 mm height in the fuel tube.

• Assume all air entering the core reacts with graphite to form CO.

The analysis yielded the following results:

• The helium up-flow is laminar and occupies 86 percent of fuel tube flow area

• The air down-flow is turbulent and occupies 14 percent of fuel tube flow area

• The volumetric up-flow = the volumetric down-flow = 0.00151 m3/s

• The air mass flow = the air ingress rate = 4.18E-04 kg/s

• The air inflow would replace the helium in the core and upper and lower plenum in 10.3 hours

• The CO formation rate = 0.0283 Mol CO/s

• The reaction energy = 3.13 KW

• The decay heat at 3 hours = 5 MW

Conclusions:

• The air ingress through a ruptured fuel tube is small.

• The CO reaction energy from air ingress through a ruptured fuel tube is significantly larger than that from air ingress due to the contraction mechanism.

• The CO reaction energy from air ingress through a ruptured fuel tube is negligible compared to the decay heat.The analysis is conservative (pure air in RB, equal temperatures, all O2 reacts to form CO).

Therefore, both air ingress mechanisms yield a negligible rate of graphite oxidation and have no impact on the RB requirements.




A3.12 CONCLUSIONS AND RECOMMENDATIONS FOR FURTHER STUDY

The analysis of dose consequences for the RB alternatives provides the following results and conclusions:

• Some mitigation by a RB is required to meet the thyroid PAG dose limit.

• All evaluated RB alternatives provide ample margin to meet 10CFR50.34 and PAGs dose criteria evaluated in this study.

• Alternative 3a provides the most effective radiological retention for DLOFC across the break spectrum for the RBVV = 10 percent cases; the doses for Alternative 3b are somewhat lower than 3a due to the increased volume and provides the best radionuclide retention capability of all the alternatives evaluated in this study.

• Alternative 2 provides comparable radiological retention to Alternative 3a and 3b for DLOFC with design basis break sizes up to 100 mm break size.

• Alternatives 1a and 1b provide an indication of the consequences of events where design features such as filters and re-closable dampers added in Alternatives 2, 3a and 3b fail (BDBEs)

• Alternatives 4a and 4b provide less effective retention than Alternatives 3a and 3b across the full break spectrum and less than Alternative 2 across the DBE spectrum because the RB is pressurized to drive out radio-nuclides for the duration of dose calculation despite a relatively low leak rate

• Alternative 4c was defined to simulate the consequences of an assumed delayed failure of the RB (a BDBE assumption). The doses from Alternative 4c exceed the dose limits for all the cases analyzed including the case with no reactor building, because the RB failureresults in a large prompt release from the pressurized RB, compared to the slower releases from a depressurized RB in the other alternatives.

• The limiting RB pressure is from the 1000 mm DBA break of the SHTS piping. The 10percent RBVV design would need to withstand a peak pressure of 5.1 bar. The peak pressure is reduced to 3.3 bar, 2.2 bar, and 1.5 bar if the RBVV is increased to 20, 50, and 100 percent of the total RB volume, respectively. In the conceptual design, strategies to improve the reliability of the SHTS HPB should be considered to reduce the reactor building design pressure and/or volume of the vented area.

• The RB pressure capacity discussed in the previous bullet would also maintain RB integrity for the BDBEs and for design alternative 4.

• The chosen RB alternative should be optimized for minimum cost by considering the tradeoffs between the design pressure, the RB vent volume, the RB vent volume leak rate, the re-closeable damper opening pressure and the filter damper opening pressure.




The conclusions regarding the radiological retention capability of the evaluated RB alternatives are subject to limitations due to:

• The lack of design maturity and an associated full scope PRA model for the NGNP.

• The need to evaluate different HPB break locations and a fuller set of licensing basis events.

• The need to consider the impact of natural convection on core temperatures during small leaks (2 to 10 mm) for break locations and configurations not addressed in this study.

• The lack of a fully integrated mechanistic source term model.

• The need to consider the quantitative failure probabilities of various design features as well as the RB structural capability to withstand loads from a full set of licensing basis events.

• The lack of a full uncertainty analysis for the source term and consequence modeling.

These limitations should be addressed in the Conceptual and Preliminary Design Stages of the NGNP.

A detailed mechanistic code with integrated PHTS and RB models, in place of the simplified 3 volume model, will be required for the next phase of the analysis.




A4 INTEGRATED EVALUATION OF REACTOR BUILDINGDESIGN CONCEPT ALTERNATIVES

A4.1 TECHNICAL APPROACH TO EVALUATION

The focus of this part of the study was on design strategies to perform pressure relief,radiological retention, and control of air-ingress functions. A total of 7 reactor building concepts was defined in Section A2 including those listed in Table 7. These concepts include variations of unfiltered vented concepts, filtered-vented concepts, and pressure retaining concepts. The parameters that were varied in the evaluation of these concepts included the leak rate for the area of the building that would be exposed to blow-down loads and fission product release pathways (Reactor Building vented area), use of rupture panels vs. open vented areas to manage the pressure relief function, use of filters to mitigate the delayed fuel release, blow-down phase of the release, or both, use of isolation devices to protect the normal operation HVAC system and to isolate the pressure relief shaft following blow-down, and alternative volumes of the vented area.

The above alternatives were evaluated using a qualitative multivariate decision analysis approach based on engineering judgment of the Westinghouse project team supported by radiological release calculations, design pressure calculations for the selected LBEs, as well as a preliminary assessment of the relative costs of the different alternatives. The evaluation criteria and weighting factors used in this evaluation are summarized in Table 17. For each criterion, the alternative ranked the best according to that criterion was given a score of 10 and the remaining alternatives given scores from 1 through 10 based on how they compared to the top ranking alternative for that criterion.

Table 17 Evaluation Criteria and Relative WeightsCriteria Relative Weight

Normal Operation Requirements 10%Investment Protection Requirement 5%Safety Functional Requirements 35%

HPB leaks/breaks 20%Seismic 5%Hydrogen/process hazards 10%

Security /aircraft crash 10%Capital and Operating Cost 25%Licensability 15%

Total 100%




A4.2 EVALUATION OF REACTOR BUILDING ALTERNATIVES

The pressure transient and radionuclide retention capability of alternatives 1a through 4b have been evaluated using a judgment based scoring system. This process has followed these basic steps:

• Each alternative is presumed to be capable of meeting all non-safety functional requirements (mostly geometry and strength)

• Approximate pressure transients are calculated for several break sizes and locations in Section A3

• Engineered features (rupture panels, filters, etc) and possible changes in building geometry and strength requirements are estimated for each alternative

• Radionuclide retention by inherent and engineered features is estimated for each alternative, and site boundary doses are estimated in Section A3

• Cases are evaluated in Section A3 in which reactor building design features successfully perform their functions as well as cases to simulate the effects of various failure modes that may be identified in a future NGNP PRA

• Capital costs are estimated for each alternative with a focus on the relative costs of the features that differentiate the alternatives

The areas in which judgments are created are normal operating requirements, investment protection requirements, safety requirements, security and aircraft crash resistance requirements, costs, and licensability. Safety performance is judged by combining judgments on the alternative’s pressure transient performance, its radionuclide retention performance, its seismic response, and its ability to withstand external hazard events related to the NGNP hydrogen process design. Cost performance is a combination of judgments related to capital cost and operating cost. Licensability is judged based on expert opinion regarding the expected difficulties that the concept may encounter in the licensing arena. An arbitrary score of 10 is applied to the alternative judged to best meet the requirements for each area. Other alternatives are given scores less than ten, based on the extent to which their ability to meet the requirements is less. The capabilities that are important when judging alternatives against requirements for normal operation are good operator access during operation and maintenance, minimum operator exposure to radiation and other hazards, and minimum normal operating release of radio-nuclides to the offsite public.




A4.2.1 Evaluation of Normal Operation Criterion

In the area of normal operation, the alternatives were judged as follows:

Alternative 1a Adequate access and control of exposure; May not provide control of air activation products, and may require large HVAC flow.

Resulting score: 9

Alternative 1b Adequate access and control of exposure; vent path normally isolated from environment improves control of air activation products and does not require large HVAC flow.

Resulting score: 10

Alternative 2 Same as 1b. Resulting score: 10

Alternative 3a Reduced leak rate requires airlocks and other impediments to operator access, otherwise, similar to 2.

Resulting score: 9

Alternative 3b Same as 3a. Resulting score: 8

Alternative 4a and 4b

Pressure retention requires additional features that impede operator access.

Resulting score: 3

A4.2.2 Evaluation of Investment Protection Criterion

In the area of investment protection, important attributes are low forced outage rate and durations, low risk of events that can cause damage to plant systems, structures, and components (SSCs), low risk of events that can result in significant plant outage time, and acceptably low risk of plant write-off. In this area, the alternatives were scored lower if they are judged to have an increase in forced outages and an increase in downtime. Those with more stringent leakage requirements were assumed to have an increase in outage durations and to take more time to recover from radionuclide releases.

Alternative 1a Easy access, but more susceptible to external factors causing outages.

Resulting score: 9

Alternative 1b Easy access, less susceptible to external factors causing outages.

Resulting score: 10

Alternative 2 Access hindered, but shorter recovery from events. Resulting score: 8

Alternative 3a Assumed to have an increase in outage durations (hindered accessibility) and to take more time to recover from radionuclide releases.

Resulting score: 6




Alternative 3b Assumed to have an increase in outage durations (hindered accessibility) and to take more time to recover from radionuclide releases.

Resulting score: 6


Assumed to have an increase in outage durations (lack of accessibility) and to take more time to recover from radionuclide releases.

Resulting score: 4

A4.2.3 Evaluation of Safety Criterion

A4.2.3.1 Evaluation of Helium Pressure Boundary Break Response

When judging alternatives for their safety response to HPB breaks, the over-arching goal is to maintain the geometry of the reactor and its passive heat rejection system (RCCS). This requires that the building be design to reliably resist the pressure transient loads.

Alternative 1a A vented design that results in survivable pressure transient with high reliability.

Resulting score: 10

Alternative 1b Similar to 1a, but has the addition of rupture panels, which increase pressure transient slightly. This addition will help to isolate and protect HVAC.

Resulting score: 9

Alternative 2 Has higher pressure transients, but with additional cost, could be designed to resist the expected load therefore scored the same as 1b.

Resulting score: 9

Alternative 3a Has higher pressure transients, but with additional cost, could be designed to resist the expected load therefore scored the same as 1b.

Resulting score: 9

Alternative 3b Has higher pressure transients, but with additional cost, could be designed to resist the expected load therefore scored the same as 1b.

Resulting score: 9


Has higher pressure transients, but with additional cost, could be designed to resist the expected load therefore scored the same as 1b.

Resulting score: 9

A4.2.3.2 Evaluation of Radiological Retention Criterion

When judging alternatives for their response to radionuclide retention, calculations have been performed that provide an estimated building response over a spectrum of postulated HPB




break events and these are summarized in Section A3. Scores for each alternative are correlated to offsite doses over these events.

Alternative 1a Given lower scores because it indicates higher doses (although still below design goals).

Resulting score: 7

Alternative 1b Given lower scores because it indicates higher doses (although still below design goals).

Resulting score: 7

Alternative 2 Given lower scores because it indicates higher doses (although still below design goals).

Resulting score: 8

Alternative 3a Receive highest scores based on the results of the radiological release study. It should be noted that has possible event sequences in which the filters do not operate.

Resulting score: 10

Alternative 3b Receive highest scores based on the results of the radiological release study. It should be noted that has possible event sequences in which the filters do not operate.

Resulting score: 10


Have a low score because some events (small helium breaks) result in the release and transport of both prompt and delayed fuel source terms driven by pent-up non-condensable helium retained in the low leakage building.

Resulting score: 5

A4.2.3.3 Evaluation of Seismic Capability

When judging alternatives for their response to seismic requirements, it is assumed that all alternatives are capable of resisting seismic forces. Design alternatives are less desirable if they require that heavy systems or components must be placed at high elevations in the building.All alternatives designed for seismic with margin and must respond with capability for beyond SSE events. Those with fewer filters and dampers are graded higher. Those with higher leak tightness more susceptible to seismic-induced cracks and leakage and are graded lower. This criterion is impacted by degree of embedment, which is discussed in later sections of the report.

Alternative 1a Fewest filters and dampers. Resulting score: 10

Alternative 1b Fewest filters and dampers. Resulting score: 10

Alternative 2 Similar to 1b but has filters and damper on top of building.

Resulting score: 9

Alternative 3a Has larger filters mounted on top of building. Resulting score: 8




Alternative 3b Has larger filters mounted on top of building. Resulting score: 8


Susceptible to seismic-induced cracks and leakage. Resulting score: 6

A4.2.4 Evaluation of Process Hazards Criterion

When judging alternatives for their response to process hazard design goals, important criteria include the ability to protect safety related components and functions from shock or pressure waves, or toxicity of chemical hazards. This area is also impacted by the degree of building embedment. The alternatives were judged as follows:

Alternative 1a External pressure loading due to hydrogen explosion could be greater than pressure transient loading; Open vent path may weaken resistance to hydrogen explosion.

Resulting score: 5

Alternative 1b External pressure loading due to hydrogen explosion could be greater than pressure transient loading; Open vent path may weaken resistance to hydrogen explosion.

Resulting score: 5

Alternative 2 Superior to open alternatives 1a and 1b. Resulting score: 8

Alternative 3a More robust building means hazards not likely to control or impact design.

Resulting score: 9

Alternative 3b More robust building means hazards not likely to control or impact design.

Resulting score: 9


Even more robust building, without vent or filters, offers the greatest resistance to hydrogen event hazards.

Resulting score: 10

A4.2.5 Evaluation of Physical Security and Aircraft Crash Criteria

When judging alternatives for their response to security threats or aircraft crash hazards,the goals are to prevent malevolent intervention from impacting plant safety or operation and to protect RB internals from being impacted by airplane crash, including fuel fires. This area may be impacted by the degree of embedment; however, the impact is probably the same for all alternatives.

Alternative 1a Judged poorer than all the other alternatives because of the open vent path.

Resulting score: 5

Alternative 1b Judged poorer than all the other alternatives because of the open vent path.

Resulting score: 5




Alternative 2 No directly open path for ingress of fluid or debris. Resulting score: 10

Alternative 3a No directly open path for ingress of fluid or debris. Resulting score: 10

Alternative 3b No directly open path for ingress of fluid or debris. Resulting score: 10


No directly open path for ingress of fluid or debris. Resulting score: 10

A4.2.6 Evaluation of Capital and Operating Costs

Alternatives are judged on estimated capital cost based on rough estimates from other project data. The effect due to operating cost is correlated with leak-tightness and number of active components. Costs are clearly impacted by the degree of embedment. In the area of cost, the alternatives were judged as follows:

Alternative 1a Minimal features and loads. Resulting score: 10

Alternative 1b Modest additional features. Resulting score: 9

Alternative 2 Additional damper and filter, not significant cost drivers. Resulting score: 8

Alternative 3a Significant cost penalty for reduced leak rate. Structural design probably not controlled by pressure loads.

Resulting score: 5

Alternative 3b Similar to 3a, but with additional cost for expansion volume.

Resulting score: 4

Alternative 4a Large cost increment for low leak rate, pressure retaining capability.

Resulting score: 2

Alternative 4b Similar to 4a, but with modest reduction in cost due to lower pressure load; additional cost for expansion volume.

Resulting score: 1

A4.2.7 Evaluation of Licensability Criterion

When judging alternatives for their response to licensability goals it is important that they meet statutory limits for LBEs at site boundary. It is a project design goal that the NGNP not require and evacuation drills. This goal means that the designs must also meet EPA protective action guidelines (PAGs) for LBEs for no evacuation at site boundary. This assessment is highly subjective, and is linked to the margins of the safety criteria and the ability to credibly present and defend a non-LWR safety design approach. Evaluation of the licensability of alternatives is as follows:




Alternative 1a Given a low licensability due to the difficulty in proving that design features such as dampers and filters are not cost effective and that the design is consistent with defense-in-depth principles.

Resulting score: 4

Alternative 1b Given a low licensability due to the difficulty in proving that design features such as dampers and filters are not cost effective and that the design is consistent with defense-in-depth principles.

Resulting score: 4

Alternative 2 Vented options are ranked high based on their dose margins.

Resulting score: 9

Alternative 3a Vented options are ranked high based on their dose margins.

Resulting score: 10

Alternative 3b Vented options are ranked high based on their dose margins.

Resulting score: 10


Have the potential for the highest consequence sequences.

Resulting score: 6

A4.2.8 Summary of Evaluation Results

The scoring of the Reactor Building alternatives against each individual criterion, without the weighting factors, is shown in Table 18, and the combined effects of the scores and the weighting factors is shown in Table 19. As seen in the results of the integrated evaluation, all of the vented options scored higher than either of the two pressure retaining options when all of the factors were considered in an integrated fashion. Alternative 2 followed by Alternative 3a were the highest ranking alternatives and should be considered for further evaluation in the Conceptual Design phase of the NGNP. Because the volume of the vented area of the Reactor Building is a parameter that must be optimized against many factors not addressed in this evaluation, Alternative 3b is also a good candidate for further evaluation. After understanding the importance of the assumed reactor building leak rates in the radiological evaluation in Section A3, the use of additional features such as reclosable dampers should also be considered across all alternates during conceptual design.

It is noted that all of the options considered in this evaluation could be applied to any level of reactor embedment. It is also expected that with full embedment, a somewhat lower leak rate of the Reactor Building vent area might be achieved, however lower leak rates could also be achieved via engineered features on the building such as seals and special doors without any special level of embedment.




Table 18 Evaluation Scores for Individual Criteria

Alternative 1a 1b 2 3a 3b 4a 4bNormal Operating Requirements 9 10 10 9 8 3 3Investment Protection Requirements 9 10 8 6 6 4 4Safety Requirements

HPB Breaks - Pressure Response 10 9 9 9 9 9 9HPB Breaks - DoseResponse 7 7 8 10 10 5 5Seismic Response 10 10 9 8 8 6 6

Hydrogen/Process Hazard Response 5 5 8 9 9 10 10Security / Aircraft Crash Response 5 5 10 10 10 10 10Capital and Operating Cost 10 9 8 5 4 2 1Licensability 4 4 9 10 10 6 6

Table 19 Total Evaluation Scores for Reactor Building Alternatives

Alternative weight 1a 1b 2 3a 3b 4a 4bNormal Operating Requirements 10 90 100 100 90 80 30 30Investment Protection Requirements 5 45 50 40 30 30 20 20Safety Requirements HPB Breaks - Pressure

Response 12 120 108 108 108 108 108 108 HPB Breaks - Dose Response 8 56 56 64 80 80 40 40

Seismic Response 5 50 50 45 40 40 30 30Hydrogen/Process Hazard Response 10 50 50 80 90 90 100 100Security / Aircraft CrashResponse 10 50 50 100 100 100 100 100Capital and Operating Cost 25 250 225 200 125 100 50 25Licensability 15 60 60 135 150 150 90 90 TOTAL 100 771 749 872 813 778 568 543




A4.2.9 Evaluation Summary of Alternative Reactor Building ConceptsIn general the vented options that were considered (1a, 1b, 2, 3a, and 3b) were found to

be superior to the pressure retaining options (4a and 4b) based on the following considerations:

• Greater compatibility with a non-condensable and inert primary coolant

• Venting of the primary coolant inventory to atmosphere with or without filtration eliminates a driving force for subsequent fission product transport of the delayed fuel release source term

• When used with filtration (2, 3a, and 3b) provides more effective retention of radio-nuclides for the design basis event spectrum up to 100 mm. Alternatives 3a and 3b provide superior retention for beyond design basis event break sizes up to 1,000 mm as well

• Lower capital and operating costs

• Easier and less costly to engineer interfaces with RCCS, SHTS, FHSS, HSS, and other NHSS and auxiliary systems

The highest rating of integrated evaluation for alternatives examined was Alt 2 (Partially filtered and vented with rupture panels) followed by Alt 3a (Fully filter and vented with rupturepanels).

• Both alternatives (2 and 3a) provide superior radionuclide retention capability for design basis HPB breaks with DLOFC than the pressure retaining alternatives (4a and 4b); Alt 3b closely followed by 3a is superior to all evaluated alternatives across the entire HPB break spectrum including AOOs, DBEs, and BDBEs.

• Alts 2, 3a, and 3b are expected to have greater licensability than either of the open vented options (1a and 1b) due to their superior capability to mitigate releases and air ingress.

• Another alternative for future study is a vented building with a passive re-closabledamper without a filter. This is expected to have delayed fuel release retention capabilities approaching that of Alt 2 due to the capability to achieve a lower leak rate.

• Results of the radionuclide retention study show that all the evaluated alternatives provide sufficient margins to offsite dose limits based on inherent and passive safety characteristics of the PBMR NGNP.

• Added engineered features such as filters and re-closable dampers add additional margins.

• This study confirms that radiological retention is not a required safety function but rather a supportive safety function for the NGNP reactor building according to how these terms are defined in the NHNP risk informed and performance based licensing approach.




Having said that it is noted that the required safety functions of the reactor building that involve the structural protection of the reactor and its inherent and passive safety characteristics also serve to maintain the fundamental safety function of controlling radionuclide releases. It supports this function primarily by keeping the radio-nuclidesinside the coated particle fuel and PHTS HPB and secondarily by retaining radio-nuclidesthat may be released from the fuel and HPB.

• In order to support the PBMR NGNP capabilities for defense-in-depth, it is recommended that a design goal be set for a radiological retention capability of a factor of 10 reduction in releases from the RB relative to that released from the HPB for I-131and Cs-137 for DBE and BDBE HPB breaks.




A5 OPEN ISSUES AND ADDITIONAL ENGINEERING STUDIES

Conclusions regarding radiological retention capability of evaluated RB options are subject to limitations due to:

• Lack of design details and associated full scope PRA model

• Need to evaluate different HPB break locations and a fuller set of licensing basis events

• Need to consider the impact of natural convection on core temperatures during small leaks (2-10 mm)

• Lack of a fully integrated mechanistic source term model

• Need to consider the failure probabilities of various design features as well as RB structural capability to withstand loads from a full set of licensing basis events

• Lack of a full uncertainty analysis in the source term and consequence modeling

These limitations should be addressed in the Conceptual and Preliminary Design Stages of the NGNP supported by the NGNP PRA

Key challenges were identified for reactor building design that need to be addressed in the conceptual design stage. These challenges include:

• Need for optimization of RV vented volume dimensions vs performance and cost

• Unknowns regarding needed protection against hydrogen process hazards

• Unknowns regarding needed protection against physical security threats

• Systems interactions issues associated with SHTS piping penetration RB walls

• Key requirement for the RB is to provide physical separation of the NHSS from events and hazards associated with the HPS, PCS, and BOP facilities

• SHTS piping provides structural linkage between RB and adjacent buildings

There is a need to investigate further the possible “systems interactions” involving:

• Faults in HPS or PCS propagating into RB

• External events for which RB is protected but other buildings are not causing adverseinteractions

• Need to provide high confidence of no adverse interactions for design basis events




• May lead to identification multiple large HPB breaks for BDBEs during the Conceptual Design PRA

• Key challenge in the next phase of the design

• Issue may complicate the approach to embedment

Pending a more thorough design iteration, DDNs will be formulated leading to potential technology development primarily in the fuel, reactor, and HPB.




PART B: EVALUATION OF REACTOR EMBEDMENT

This portion of the study develops requirements and criteria for determining the degree of embedment of the reactor. This study of the PBMR reactor considers the interaction among factors that influence the depth of the embedment. These factors include cost, design basis threats, seismic effects, and hazards resistance. The results of this study will be used to characterize the interactions of these factors on embedment depths for commercial application of this technology. References from relevant sections of the Electric Power Research Institute (EPRI) Advanced Light Water Reactor Utility Requirements Document are reviewed for applicability in this study description.

In addition, this study evaluates the potential differences embedment could make fortransferring heat to the environment including heat flow through walls to ground or to the air during severe beyond accident conditions.

This study assesses contributing features with respect to siting of the NGNP at INL and a commercial facility sited at other potential locations within the US under the jurisdiction of the US NRC.

This portion of the study is conducted as indicated in the following flow chart (Figure 24).




Figure 24 Flow Chart of Part B: Evaluation of Reactor Embedment




B1 LIST OF ISSUES RELEVANT TO REACTOR EMBEDMENT

The following factors are considered in this section of the report. Reference information provided in Ref. [8] has been reviewed along with other regulatory sources and discussed as required.

• Operational Needs including equipment layout

• Heat Dissipation to Environment

• Water Table Effects

• Geotechnical Constraints and Foundation Performance

• Construction Considerations

• Cost Consideration

• Malevolent Hazards

• Natural Phenomenon Hazards

• Natural Geological Hazards

• Chemical Releases, Explosions, and Manmade Hazards

B1.1 OPERATIONAL NEEDS

The Reactor and the reactor building layout should consider the following with respect to embedment for safe reliable and efficient operation and maintenance of the Nuclear Heat Supply Facility.

• Maintenance access must be provided to all major equipment.

• Safe personnel access and egress paths (elevators, corridors and stairwells) must be provided.

• Enable placement of system components as needed relative to the elevation of the reactor itself, to meet thermal hydraulic and other operational needs.

• Protection must be provided of safety related components from external hazards and hazards due to internal SSC failures.

• The PBMR design does not require routine access to the reactor for refueling activities. (The need to access the reactor head for refueling drives other HTGR configurations to favor full embedment.)

• The Pressure Relief System (PRS) is anticipated to discharge from a vent stack located on the top of the building. This need could benefit with a building roof elevated above




ground. It is not anticipated that an elevated release point requiring a tall stack will be needed to meet off site dose limits. Analysis in the section A assumes ground level releases for all cases. Therefore there is no particular benefit with a taller building.

• Optimized travel distances for routine operator access and egress. Travel distances can have an impact on life cycle costs as discussed in section B1.6.

The following information has been provided by the PBMR design team and reflects current work in progress in development of the reactor building layout for a single module design for a Brayton Cycle Electrical Machine (the DPP) and is provided for discussion purposes only. It is intended that during the conceptual design phase of the NGNP a reactor building layout will be developed using the functional requirements identified in this study report as a basis (see part A). The degree of embedment does not significantly affect plant operations. The Reactor building layout currently being considered includes access to the Reactor Building from the Auxiliary Building. This is either via a main equipment hoist lobby or at the lowest level in the building. However, the access to the reactor building via the main equipment elevator lobby can be at any level. Each level has a lobby and direct access to the elevator. It is expected the equipment maintenance elevators and hoist spaces to transport major components to these access points can be readily configured regardless of level of embedment.

Another key feature that leads to the conclusion that the level of embedment is not significant to the design for operations is that the refueling system uses continuous circulation of the fuel spheres and does not require routine overhead access to the reactor as is the case with other HTGR design using prismatic block core structures. This is the main reason why other HTGR designs have gone to the embedded configuration.




Figure 25 Access to Reactor Building




Figure 26 Access to Reactor Building at Main Equipment Elevator

Figure 27 Access to Reactor Building at Lowest Level in the Building




Figure 28 Elevation Requirements of Major SSC With Respect to the Reactor

The following key systems have specific location requirements with respect to the Reactor:

• The Reactivity Control System, Reserve Shutdown System and Core Unloading Device are an integral part of the Reactor Unit Assembly.

• The Fuel Handling and Storage System (FHSS) valves and piping.

• The Reactor Cavity Cooling System storage tanks must be higher than the Reactor to facilitate the passive cooling function. The RCCS stand pipes must be at the same elevation as the RPV within the reactor cavity.




• The In Core Delivery System must be above the reactor to facilitate top entry into the core.

• The Core Conditioning System (CCS) is located below the Core Outlet Pipe (COP).Positioning of the CCS heat exchanger circuit should be below the Reactor due to reduce the probability of water leakage into the core. The CCS is located as close to the COP as possible to reduce the Hot Pipe length.

• The HSS Tanks and Fuel Storage Tanks are located as low in the building as possible due to their size and weight. In the DPP an advantage of placing them below ground level is the additional protection from external events.

Table 20 identifies the major systems to be housed in the Reactor Building, the relative safety classification and requirements for location with respect to the Reactor and other systems.The following safety classifications were taken from Reference [2] for input to this table. The safety classification are assumed for the purposes of this study only in order to gain an understanding of the portion of the Reactor building that will house SSCs requiring protection from internal and external hazards. Table 20is not to be interpreted as a formal position on the SSC safety classification. Safety classification will be developed during conceptual and preliminary design based on the RI-PB safety analysis and licensing approach.

Safety-Related SSCs (SR):

This category is for SSCs relied on to perform required safety functions to mitigate the public consequences of Design Basis Events (DBEs) to comply with the dose limits of 10 CFR §50.34[8]

This category is also for SSCs relied on to perform required safety functions to prevent the frequency of Beyond Design Basis Accidents (BDBEs) with consequences greater than the 10 CFR §50.34[8] dose limits from increasing into the DBE region.

Non-Safety-Related with Special Treatment (NSR-ST):

This category is for SSCs relied on to perform safety functions to mitigate the consequences of Anticipated Operational Occurrences (AOOs) to comply with the offsite dose limits of 10 CFR Part 20[79]

This category is also for SSCs relied on to perform safety functions to prevent the frequency of DBEs with consequences greater than the 10 CFR Part 20 [79] offsite dose limits from increasing into the AOO region.

Non-Safety-Related with No Special Treatment (NSR)

This category is for all SSCs not included in either of the above two categories.




The basic conclusion drawn from Table 20 is that the required reactor building foot print is likely to be larger than just the reactor cavity and citadel structure at all elevations including the lower elevation and therefore the foot-print to be embedded is quite large. Figure 29 below provides the initial layout of the NGNP under consideration which is based on the NGNP PCDR baseline.

ReactorBuildingReactorBuilding

Figure 29 NGNP Reactor Building Layout

NG

NP-

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tem

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NO

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Loca

tion

clos

e to

bot

tom

of R

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or

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el is

des

irabl

e to

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imiz

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lium

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t Exc

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tion

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or V

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l pre

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s po

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le g

ravi

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ater

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age

to re

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oC

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Pipi

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NO

SR

Arra

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an b

e op

timiz

ed to

sui

t em

bedm

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onst

rain

ed b

y C

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and

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requ

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Loca

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for N

atur

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val s

tand

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es o

n sa

me

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PR

V to

pro

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reac

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avity

con

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e.

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age

Syst

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Sto

rage

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syst

emN

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SR-S

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oSp

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Circ

ulat

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Subs

yste

mYE

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rabl

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mai

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n D

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eom

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Rep

leni

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ubsy

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YES

NSR

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Des

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deve

lopm

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xten

sive

It is

de

sira

ble

to m

aint

ain

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oFu

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NO

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Arra

ngem

ent c

an b

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timiz

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sui

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bedm

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g G

as S

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rang

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t can

be

optim

ized

to s

uit

embe

dmen

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oSp

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Dec

omm

issi

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bsys

tem

NO

NSR

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Arra

ngem

ent c

an b

e op

timiz

ed to

sui

t em

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ent

oAu

xilia

ry G

as S

ubsy

stem

NO

NSR

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Arra

ngem

ent c

an b

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timiz

ed to

sui

t em

bedm

ent

o H

igh-

leve

l Was

te H

andl

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em

NO

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Arra

ngem

ent c

an b

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timiz

ed to

sui

t em

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eliu

m S

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em

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ory

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mN

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rang

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t can

be

optim

ized

to s

uit

embe

dmen

t

NG

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NO

NSR

-ST

Arra

ngem

ent c

an b

e op

timiz

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sui

t em

bedm

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oH

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optim

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to s

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Con

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Syst

em

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SR-S

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NO

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NO

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Arra

ngem

ent c

an b

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timiz

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t em

bedm

ent

•N

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Coo

ling

Wat

er S

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oAu

xilia

ry C

ompo

nent

Coo

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Wat

er S

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SR-S

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rang

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t can

be

optim

ized

to s

uit

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dmen

t

oEq

uipm

ent P

rote

ctio

n C

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g C

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t N

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rang

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t can

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optim

ized

to s

uit

embe

dmen

t

•N

HSS

Ele

ctric

al S

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mN

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rang

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t can

be

optim

ized

to s

uit

embe

dmen

t

•N

HSS

HVA

C S

yste

mN

ON

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rang

emen

t can

be

optim

ized

to s

uit

embe

dmen

t

•Pr

imar

y Lo

op In

itial

Cle

anup

Sys

tem

NO

NSR

-ST

Arra

ngem

ent c

an b

e op

timiz

ed to

sui

t em

bedm

ent

•N

HSB

Pres

sure

Rel

ief S

yste

mN

OSR

Arra

ngem

ent c

an b

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timiz

ed to

sui

t

NG

NP-

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S 10

0-R

XB

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0N

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once

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y Rea

ctor

Bui

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nctio

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nd T

echn

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rem

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and

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of R

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•H

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rans

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Syst

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eat T

rans

port

Syst

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•PH

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NO

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Nee

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alve

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term

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X C

ore

NO

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To b

e de

term

ined

dur

ing

Con

cept

ual

Des

ign.

•In

tern

al D

ucts

, Sup

ports

and

Insu

latio

nN

ON

SR-S

TTo

be

dete

rmin

ed d

urin

g C

once

ptua

l D

esig

n.

•IH

X Ve

ssel

, Inc

ludi

ng E

xter

nal S

uppo

rts a

nd

Insu

latio

nN

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To b

e de

term

ined

dur

ing

Con

cept

ual

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ign.

•Pi

ping

(prim

ary

circ

uit f

rom

Rea

ctor

to IH

X an

d be

twee

n IH

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ssel

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oth

hot a

nd c

old

legs

)

•Pr

essu

re B

ound

ary

Pipi

ng, I

nclu

ding

Ex

tern

al S

uppo

rts a

nd In

sula

tion

YES

SRN

eed

to m

inim

ize

Hel

ium

Pip

ing

Leng

ths

•Pi

ping

Inte

rnal

Duc

ts, S

uppo

rts a

nd

Insu

latio

nYE

SN

SR-S

TN

eed

to m

inim

ize

Hel

ium

Pip

ing

Leng

ths

•PH

TSPr

essu

re R

elie

f Sys

tem

NO

SR

oSe

cond

ary

Hea

t Tra

nspo

rt Sy

stem

(SH

TS)

NG

NP-

NH

S 10

0-R

XB

LDG

Rev

0N

GN

P C

once

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l Des

ign

Stud

y Rea

ctor

Bui

ldin

g Fu

nctio

nal a

nd T

echn

ical

R

equi

rem

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and

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luat

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of R

eact

or E

mbe

dmen

t

NG

NP-

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S10

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200

812

5 of

186

Nuc

lear

Hea

t Sup

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Syst

em (N

HSS

)Fi

xed

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R

PV

NG

NP

Safe

tyC

lass

Rem

arks

•SH

TS C

ircul

ator

NO

NSR

-ST

Nee

d to

min

imiz

e H

eliu

m P

ipin

g Le

ngth

s

•H

eliu

m Is

olat

ion

Valv

es (i

f req

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SR-S

TN

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to m

inim

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Hel

ium

Pip

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Leng

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(sec

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t, ho

t and

col

d le

gs, p

lus

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SG

)N

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inim

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Hel

ium

Pip

ing

Leng

ths

•Pr

essu

re B

ound

ary

Pipi

ng, I

nclu

ding

Ex

tern

al S

uppo

rts a

nd In

sula

tion

NO

NSR

-ST

Nee

d to

min

imiz

e H

eliu

m P

ipin

g Le

ngth

s

•Pi

ping

Inte

rnal

Duc

ts, S

uppo

rts a

nd

Insu

latio

nN

ON

SR-S

TN

eed

to m

inim

ize

Hel

ium

Pip

ing

Leng

ths

•SH

TS F

low

Cou

plin

g an

d M

ixer

NO

NSR

-ST

Nee

d to

min

imiz

e H

eliu

m P

ipin

g Le

ngth

s

•SH

TS P

ress

ure

Rel

ief S

yste

mN

OSR




B1.2 HEAT DISSIPATION TO THE ENVIRONMENT

The ability of the building to transfer heat from the reactor through the reactor cavity to the environment either to ground or to ambient air is considered. First it is important to understand that the transfer of heat from the reactor is normally through the Heat Transport System (HTS) to the Power Conversion System (PCS), with a small portion of waste heat going to the RCCS. When the HTS and PCS are not available the CCS is used to remove heat. In a design basis event where the CCS is no longer available, the RCCS would be the heat transfer pathway to control the temperature of the reactor cavity. Heat transfer through the reactor cavity wall to the environment would only occur in a very unlikely beyond design basis event where the RCCS is emptied of water and no longer effective to remove heat.

Prior work by PBMR has led the team to expect that the thermal capacity of the Reactor Citadel walls is so high that the external medium will not influence the internal concretetemperatures significantly.

As an additional check, PBMR has completed a scoping analysis and documented it inreference [28]. The basis for this analysis was a BDBE which combined a DLOFC and a loss of RCCS water on initiation of the event. The objective of this calculation is to determine if there is a significant difference between heat transfer from the reactor citadel to air, soil or clay. The analysis considers heat transfer to air, sandy soil (low moisture content and low thermal conductivity) and clay (high moisture content and high thermal conductivity) with a high water content. Figure 30 shows a schematic view of this heat transfer model of the reactor cavity.

The maximum and average concrete temperatures at a selected height are plotted below in Figure 31 and Figure 32 to see the difference for the various media. The results also include a reference calculation with a fixed boundary condition of 40ºC on the outer surface of the reactor cavity. The maximum concrete temperature is the inside temperature, while the average is a mathematical average across the width but at a specific height. It can be seen that the maximum temperature is almost identical, while the average temperature only differs slightly. The average concrete temperature is only slightly cooler with clay surrounding the cavity. Clay has a high moisture content and a resultant high thermal conductivity. Sand surrounding the cavity does not show an improvement above air.

It must be stressed that the transient results shown the figures are not physically realistic, as the RPV and concrete temperatures exceed their design values early on in the transient and the embedment solution will not influence the ability of the citadel to perform its function of structural support to the reactor (NHSB-2.2.1).

The conclusion from this is that heat transfer should not be a driver in the decision to embed the reactor.




Figure 30 Top View of Reactor Cavity Schematically Showing the Added Layers

Figure 31 Maximum Concrete Temperature at a Height Corresponding to theMiddle of the Pebble Bed




Figure 32 Average Concrete Temperature at a Height Corresponding to theMiddle of the Pebble Bed

B1.3WATER TABLE EFFECTS

Excluding arid regions, the design water table at most sites will likely be less than 15 m below the ground surface.

A structure founded below groundwater level must be designed to resist flotation, and to withstand static water pressures on the submerged portion of the exterior. Additionally, the submerged portion of the exterior must be waterproofed to prevent seepage into the structure.

During construction, the groundwater level must be temporarily lowered to permit construction in the dry.

Groundwater withdrawal permits must be obtained before groundwater can be removed and disposed. This requires making estimates of groundwater withdrawal rates and of potential effects on existing groundwater users.

At INL, the approximate depth to groundwater ranges from 60 m in the northern part to 275 m in the southern part. Therefore, groundwater would likely only be a design consideration in the northern part of INL, and then only for a structure that is fully-embedded or nearly fully-embedded.




B1.3.1 FlotationThe hydrostatic uplift force on a submerged body equals the volume of water displaced

times the unit weight of the displaced water. For a structure having a regular exterior cross-section, the design hydrostatic uplift force increases in direct proportion to the depth of the structure below the design groundwater level.

Typically, resistance to flotation is provided by the weight of the structure and the weight of soil overlying any extension of the foundation beyond the exterior wall. Although additional resistance could be provided by anchoring the foundation into the underlying soil or rock, for a nuclear plant the anchors would have to be designed as a safety-related system. We are not aware of any such system at a nuclear facility.

Hydrostatic uplift reduces the normal force between the base of the structure and the underlying material, reducing the sliding resistance at the base. This is usually not a significant design issue for an embedded structure, because significant resistance to sliding is developed as the side of the structure moves against the adjacent soil.

B1.3.2 Water Pressure on Embedded WallsThe submerged portion of a structure must be designed for static lateral water pressure,

which at a given depth Y below the design groundwater level, equals:

s w = ? w Y

Where: s w = static water pressure ? w = unit weight of water Y = depth below the design groundwater level

The static water pressure diagram has a linear distribution, with zero pressure at the design groundwater level and a maximum value at the base of the structure.

B1.3.3 Permanent Dewatering Systems

If it were necessary to reduce subsurface water loads on the structure, the ground water level could be lowered using a permanent dewatering system. The system might consist of pumped wells located outside the footprint of the structure, and associated discharge piping.Wells can be effective if the underlying soils are permeable, e.g. sand or gravel.

Alternatively, a zone of freely draining material could be placed beneath the base of the structure. Perforated collection pipes in the drainage layer would conduct the water to sumps, from which it would be pumped through discharge piping.




In either case, monitoring wells would be required to verify that the groundwater level is maintained at or below the design elevation.

Problems with using a permanent dewatering system for the NGNP project are:

• The system would have to be safety-related. This would likely require redundant pumps and power supply, and seismically designed discharge piping. This would substantially increase the cost of designing, constructing and operating the system.

• The potential for water sources other than groundwater would need to be considered, e.g.failure of water piping in the area.

• It might be difficult to demonstrate conclusively that the freely draining material beneath the structure would not become clogged during the life of the system, due to factors such as soil infiltration, mineral deposition or bacterial growth.

B1.3.4 Waterproofing

Waterproofing will be required on the outside of the submerged portion of the structure.Basic types of waterproofing systems are (Postma and Walker, 2006):

• Cementitious systems: These contain Portland cement with water and sand combined with an active waterproofing agent. These systems include metallic, crystalline, chemical additive and acrylic modified systems.

• Fluid-applied systems: These include urethanes, rubbers, plastics, and modified asphalts.Fluid membranes are applied as a liquid and cure to form a monolithic seamless sheet.

• Sheet-membrane systems: These include thermoplastics, vulcanized rubbers, and rubberized asphalts.

• Bentonite clays: Natural clay known as bentonite acts as waterproofing by swelling when exposed to water, thus becoming impervious to water. The bentonite is sandwiched between two layers of geo-textile or paper, and is furnished in panels or sheets.

For this study, fluid-applied systems will be considered, because they are easy to construct. They can be applied to vertical and horizontal surfaces, and can be installed in relatively tight quarters, such as in a narrow annular space around a foundation wall. .

B1.3.5 Fluid-Applied Waterproofing Membrane

A fluid-applied membrane is sprayed or rolled onto the surface to be waterproofed, and later solidifies. The fluid may be applied in one or more layers to form a membrane having a required dry thickness. After the fluid is applied the covered area is inspected for proper




thickness, and for pinholes, blisters or other voids in the membrane. If a defect is detected, additional material is applied over the defect and surrounding area until a monolithic layer of the specified minimum thickness is formed.

An example of a fluid-applied waterproofing system is the water-borne asphalt emulsion manufactured by CETCO Liquid Boot Company. This product is also used as a gas vapor membrane.

To waterproof the bottom of the structural mat, a 75 mm thick mud mat would be placed and allowed to set. A base geotextile would be placed over the mud mat, followed by the asphalt emulsion membrane (e.g., 2 mm dry thickness). After the membrane has cured and beenchecked for proper thickness and absence of flaws, a protection board or protection mat would be placed over the membrane. This would be followed by another 75 mm mud mat, and then by the structural mat. Membrane installation on vertical surfaces (mat and embedded foundation wall) would be similar, except that no mud mats would be placed.

B1.3.6 Construction Dewatering

For excavations below the groundwater table, the groundwater level must be lowered some distance (say 1 m) below the bottom of the excavation, in order to provide a reasonably firm working surface. In sands or gravels, deep wells would likely be used where a wide excavation must be made or where the depth of excavation below the groundwater table is more than 9 to 12 m, or where artesian pressure in a deep aquifer beneath an excavation must be reduced. The required rate of groundwater withdrawal, and therefore the number and spacing of wells, depends on the depth that the water surface has to be lowered, the area of the excavation to be dewatered, the permeability of soil, the presence of nearby sources of recharge (such as a stream), and the distance from the wells to the excavation.

In fine-grained silts with low permeability, gravity alone will not drain the soil, because capillary forces hold the water in the soil voids. In such case, a vacuum dewatering system can be used. The system consists of wells with the screen and riser pipe surrounded with a free-draining sand filter extending to within a few feet of the surface. The remainder of the hole is filled with impervious soil. By maintaining a vacuum in the well screen and filter, the hydraulic gradient producing flow toward the well is increased. In order to dewater this type of soil properly, it us usually necessary to install the wells fairly close together.

Preliminary estimates of soil permeability can be made from correlations with the soil grain size, and results of small-scale field permeability tests in boreholes. Such estimates can be used to identify suitable dewatering options. A field pumping test is typically performed to provide values for final design of the system.

Lowering the groundwater level at the excavation lowers the groundwater level in the surrounding area. This effect diminishes with increasing distance from the dewatering wells, and




eventually disappears where the distance is sufficiently great. Effects on existing groundwater users within the zone of influence of the dewatering system have to be evaluated, along with potential effects on existing structures within this zone. The latter is necessary because lowering the groundwater level increases the vertical effective stress on the soil. If a structure within the affected area is underlain by a layer of highly compressible soil, lowering the water level could lead to unacceptable settlement of the structure. If such a situation were identified, consideration would be given to recharging water to reduce or prevent lowering of the water table at the structure.

Groundwater withdrawal must start some time before the excavation reaches the normal groundwater level, in order to maintain the groundwater the required distance below the current bottom of the excavation.

A permit is typically required to remove groundwater during construction. Also, the quality of the water to be discharged must meet regulatory requirements.

Dewatering is typically performed by a specialty contractor. Dewatering can be costly, because the dewatering system must be operated continuously (24 hours per day, 7 days perweek). Three shifts of operators will likely be required.

.

B1.4 GEOTECHNICAL CONSTRAINTS AND FOUNDATION PERFORMANCE

B1.4.1 Temporary Excavation Support

B1.4.1.1 Excavations in Soil

Open cuts with sloped sides are feasible to very large depths, provided there is sufficient space to accommodate the wide excavation and provided the excavation has an acceptable cost.Typical excavation slopes in sand range from approximately 1.75 H: 1V to 2H: 1V. Flatter slopes are required in weaker soil. Equipment access to the excavation is by ramp having a maximum slope of approximately 10 percent. With an open cut, the top of slope is some distance from the structure to be constructed, requiring longer crane reach.

Vertical cuts require various types of excavation support, depending on the excavation depth. For shallow excavations in soil, a cantilever wall of steel sheet piles or pipe piles can be used, or a soil nail wall can be constructed. Soil nailing is a method of reinforcing existing soil by installing threaded steel bars into the cut slope as wall construction proceeds from top down.The bars are grouted in place to create a stable mass of earth. The excavated face is then covered by shotcrete, which is reinforced with wire mesh and attached to each bar by a plate.




For deep excavations, a braced secant pile wall or braced concrete diaphragm wall can be used. Bracing can be external or internal. External bracing consists of earth or rock anchors attached to horizontal beams at vertical intervals along the inside (structure side) of the wall.Internal bracing for a circular wall consists of horizontal beams at vertical intervals, with the beams acting as compression rings.

Factors affecting the design of the bracing system include:

• Soil strength and stiffness (affect the lateral soil stress on wall)

• Groundwater level during construction (determines water pressure on wall)

• Adjacent surcharge loads, such as heavy cranes

• Construction method and sequence

Note that deep cuts may cause some lateral movement of the adjacent soil. It is necessary to evaluate the potential effect on any adjacent structures or underground utilities.

B1.4.1.2 Excavations in Rock

Excavation in rock would be performed using drill and blast. Controlled blasting wouldbe performed to limit damage to the rock and to minimize breakage beyond the excavation contract pay limit. Vibration monitoring would be performed during blasting to verify that design velocities are not exceeded. Loose rock would be removed from the rock surface. The exposed rock surface would be mapped geologically to verify that no active fault passes through the excavation.

Anchor bolting would be required to prevent sliding or toppling of rock into the excavation. In addition, a protective layer of shotcrete would be placed over the rock surface.The shotcrete would be applied over welded wire fabric that is attached to the rock by short rock anchors or dowels.

• Major factors affecting design of the rock slopes include:

• Rock strength. There are more problems with weak rock.

• Orientation of discontinuities in the rock. There are more problems where slope discontinuities such as joints and shear planes slope downward into the excavation.

• Spacing of discontinuities. More support is needed for close spacing

• Shear strength along discontinuities. There are more problems with low strength.




• Presence of water in the discontinuities. Water pressure adds to the force to be resisted by the rock support system.

• The location and thickness of major soil beds between basalt flows at INL. The presence of a thick soil bed within the excavation depth increases requirements for excavation support.

B1.4.2 Lateral Earth Pressures

B1.4.2.1 Static Lateral Earth Pressure

For deeply embedded structures with rigid walls, the static earth pressures at any depth equals the at-rest lateral earth pressure, which is calculated as:

s h = Ko s ’v = Ko [(? moist) (h moist) + (? sub) (h sub)]

Where: Ko = at-rest lateral earth pressure coefficient of the soil

? moist = moist unit weight of soil

h moist = thickness of moist soil above the depth in question

? sub = submerged unit weight of soil h sub = thickness of submerged soil above the depth in question

In general, the static lateral earth pressure increases with the depth to the groundwater level. It increases at a slower rate below the groundwater table because the submerged unit weight of the soil is less than the moist unit weight..

The lateral earth pressure is increased by surcharge loading near the excavation, with the effect determined by the geometry and intensity of the loading. For surcharge of great extent, the surcharge pressure is generally taken as Ko q, where q is the applied surcharge pressure.

B1.4.2.2 Dynamic Lateral Earth Pressure

Dynamic lateral earth pressures on embedded structures are determined by soil-structureinteraction (SSI) analysis, taking into account the following:

• Relative motion between the ground and the embedded portion of the structure

• Modulus of lateral subgrade reaction, which depends on the stiffness of the structure wall and of the ground.




B1.4.3 Backfill Around Embedded Structure

Where there is room, fill typically consists of structural fill compacted to 95% of ASTM D1557 maximum dry density. Vibratory compactors are used, with the lift thickness and number of compactor coverages per lift depending on the compactor used. Light compactors are used near walls. Backfill height is kept relatively uniform around the structure to prevent unbalanced loading on the structure.

Where tight access makes placing compacted backfill difficult or uneconomical, controlled density fill can be used. This is a fluid mix of cement, sand, fly ash and water that flows easily into place, and then hardens to form a material that is as stiff as or stiffer than compacted granular fill. Controlled density fill needs no compaction. A layer is placed and allowed to set before another layer is placed.

B1.4.4 Allowable Bearing Pressure

The allowable bearing pressure is governed by the lower of the following: the pressure that has an adequate factor of safety against bearing capacity failure, and the pressure that causes the maximum allowable settlement.

Bearing capacity is generally not a significant concern for nuclear structures founded on large mat foundations, because such structures are typically founded on competent soil or on rock. Embedment increases the downward soil pressure adjacent to the base of the structure.This pressure resists a bearing capacity failure. Therefore, increasing the embedment increases the allowable bearing pressure.

Settlement estimates are performed to estimate the amount that the mat-supportedstructure will settle. In general, allowable settlement of mat-supported structures is in the order of 2 inches.

B1.4.5 Modulus of Subgrade ReactionThe modulus of vertical subgrade reaction Kv is used to characterize the stiffness of the

supporting medium in designing mats and slabs on grade using the Winkler method. Kv is defined as:

Kv = q / d

Where: q = applied stress at the soil surface

d = settlement

K v is expressed in units of (F/L2)/L which is the same as F/L3.




The magnitude of Kv depends on the stiffness of the soil, the width, shape and depth of the loaded area, and the position on the mat or slab. Therefore, Kv is not a fundamental soil property.

If all other conditions are the same, Kv increases with embedment depth, because settlements are smaller for the same applied stress. This occurs because the change in stress in the soil due to the applied q is a smaller percentage of the initial stress (Coduto, 2001).However, where the soils are competent, embedment would have only a small impact on structural design.

B1.4.6 Sliding Stability

The factor of safety against sliding is the sum of the forces resisting sliding divided by the sum of the forces causing sliding.

The resisting forces are the base shear resistance and the passive soil pressure that develops as the structure is driven into the soil on the side opposite the driving force. The latter depends on the amount that the structure moves into the surrounding soil.

The base shear resistance equals the normal force times the coefficient of sliding friction.The latter value is typically controlled by the coefficient of friction of the waterproofing system, which is dependent on the membrane material, its hardness and/or surface roughness, the type of material on either side of the membrane, the applied normal stress, the rate of loading, and whether the materials are dry or saturated. Such values are determined by shear box tests performed according to ASTM D 5321, using project specific materials and conditions.

Hydrostatic uplift reduces the normal force between the base of the structure and the underlying material, reducing the sliding resistance at the base. This is usually not a significant design issue for an embedded structure, because significant passive resistance is developed as the side of the structure moves against the adjacent soil.

B1.4.7 Structure Overturning

Wind or seismic forces exert lateral forces on a structure, creating an overturning moment. Resisting moments are developed by the weight of the structure (minus the force due to any hydrostatic pressure on the base of the mat) and by passive earth pressure acting on the side of the structure that moves against the adjacent soil. Therefore, embedment increases the factor of safety against overturning.




B1.4.8 Structure Settlement

Embedding a structure decreases the amount of settlement that the structure will undergo.This is due to the fact that the net vertical stress imposed on the underlying material equals the average bearing stress minus the effective stress that was exerted by the excavated material. This net vertical stress is always less than that imposed by a structure founded at grade.

B1.4.9 Input to Soil-Structure Interaction Analysis

Strain-compatible values of shear modulus and damping for the subsurface materials at the site are required for input to analysis of soil-structure interaction. To develop these values, borings are drilled at the site to determine the site stratigraphy. Cross-hole and/or down-holeseismic velocity surveys are performed in selected boreholes. These surveys provide a profile of shear wave velocity vs. depth for the soil and any rock adjacent to the boreholes in which the survey was made. These results are used to generate profiles of low-strain shear modulus vs. depth. Relationships for shear modulus vs. shear strain, and damping vs. shear strain, are established by laboratory test, or are adopted from the literature.

The shear modulus profiles are analyzed for response to shaking by a suite of time-histories generated by earthquakes of magnitude similar to the design earthquake, that were recorded on bedrock, and that have been scaled to the design bedrock acceleration value. The site response analysis is performed using computer codes such as ProShake (EduPro, 1999). The analyses provide strain-compatible values of shear modulus and damping for each profile.Values corresponding to the lower and upper bound profiles are then used in analyses of soil-structure interaction. These analyses apply the input motion at the base of the structure.

The process described above would be the same for an embedded structure and for a structure founded at the ground surface. The borings for the embedded structure would likely extend to somewhat greater depth, increasing the boring cost.

As discussed in Section B1.9, embedding a structure decreases the input motion at the base of the structure. Certain limits apply. For example, NRC Standard Review Plan 3.7.2 requires the horizontal component of the acceleration at foundation depth to be no less than60 percent of that at finished grade in the free field.




B1.5 CONSTRUCTION CONSIDERATIONS

Construction complexity will be assessed. The effect of the embedment depth on constructability using modern techniques such as modularization is discussed. The distance below grade at which a structure is located has a significant impact on the complexity of the construction process. It affects not only the construction of the foundations but the superstructure and interior components as well. Geotechnical conditions vary greatly from site to site and range from significant depths of soils to solid rock at or near the surface. Each geotechnical condition presents its own unique challenges when constructing the foundation.

To embed a structure on a site where solid rock is at or near the surface would start with drilling and blasting so that the rock can be removed to the depth of the bottom of the foundation. This is a costly and dangerous process that can take a long time to complete. In addition, there is potential for collateral damage to existing structures nearby due to vibrations and shock waves set up by the blasting. While the likelihood of significant flooding as the result of ground water in solid rock is small, pumping systems are still required to remove water that may seep in through cracks or that is deposited by rains.

On soil sites, the major issues include dealing with ground water and stability of the soils.The techniques for dealing with ground water generally involve some means of pumping water away from the excavation. The various dewatering techniques generally involve continuous pumping and are well proven. Disposal of the water can be a costly process if there isn’t a place to receive the water in the immediate vicinity.

In general, protection of earthen side walls from cave-in must be addressed for any excavation deeper than about 1 meter. The US Department of Labor regulations found in 29 CFR 1926 (OSHA) require that side walls be benched or sloped at 1.5 (horizontal) to 1 (vertical),unless engineered protective means are employed such as sheet piling or other bank stability or bank restraint systems.

Deep embedment of a structure without sheet piling or a similar restraint system would require that significant amounts of soil be removed and stockpiled. For a deep excavation, the amount of soil to be excavated and returned to provide slope stability can easily exceed the volume of soil required to be excavated to contain the structure. Additional amounts of soil would also have to be removed and returned for the roads necessary to drive heavy equipment into the excavation.

Depth of embedment can have a significant affect on the ability to use certain modern construction techniques such as modularization. Modularization entails assembling smaller plant components including interconnecting piping and electrical services onto a larger assembly or module at a shop located away from the congested construction area then installing the completed assembly in the plant. This reduces the installation time required at the congested jobsite.




With a deeply embedded structure, modules in the lower elevations would have to be installed very early and generally before floors are installed. This requires modules to be designed, constructed and delivered to the jobsite very early in the construction schedule which may not be practical. This also requires that the modules and their components be protected from significant hazards including weather and construction operations above including concrete placement since these modules would have to be placed in the structure before concrete floors are installed. It should be noted however that modules can incorporate structural elements including leave-in-place steel forms which can reduce the amounts of rebar that is required.

Even without modularization, deep embedment will pose certain construction challenges.Working below grade requires significant ventilation and continuous air monitoring to assure worker safety. Some welding processes require the use of inert gasses such as argon for a shielding medium. Argon gas is extremely dangerous in confined spaces since it displaces oxygen. Argon is also heavier than air and tends to find its way to the lowest spaces in the building volume thus presenting a potential hazard to workers in those areas. Above ground structures can be use construction openings to promote ventilation, whereas below ground installations do not have this option.

B1.6 COST CONSIDERATION

B1.6.1 Capital Cost

A cost analysis which included key elements of relative capital costs for the reactor building was performed [69]. The reactor building is assumed to be 65.8 m high measured from the top of concrete mat to the roof. Two assumed configurations of structures with approximately equal volumes were considered.

1. A circular structure with a 56 m outside diameter.

2. A square structure with 50 m long exterior sides

Two foundation conditions were assumed to bound the potential site geologicalconditions. The first condition assumed a rock site with a water table below the bottom of the reactor building mat under full embedment of the structure. This condition is comparable to the condition expected at the INL site. The second condition is a deep competent soil site with a high water table. This condition would be anticipated along the southeastern coast and Gulf coast of the United States.

The cost benefit analysis was based on the following assumptions:

1. Internal pressures from pipe ruptures would not govern the thickness of the exterior wall of the structure.

2. Plant layout fits within the confines of the assumed building sizes.




3. Reactor Building internals will remain essentially the same regardless of the embedment depth.

4. Below grade structures would include a concrete exterior

The key elements considered in the cost analysis [69]were the thickness of the exterior wall above and below grade, the cost of excavation, including excavation support or stabilization systems, and backfill. Use of dewatering systems was also included for soil sites.

Both structure configurations were assumed to have 1.5 m thick walls above ground to protect against malevolent hazards. While this thickness is somewhat in excess of current nuclear plant designs, the current nuclear plant designs look at these cases as a “beyond design basis event” as the result of an independent pressure boundary or containment inside of the exterior wall. The NGNP reactor building is still in the schematic design phase and might nothave a pressure boundary or containment. If it does have a pressure boundary, the pressureboundary will likely be present in a localized area of the reactor building. The use of a potentially thicker above ground exterior wall will result in favoring a deeper embedment. The assumed wall thickness is anticipated to be sufficient to cover all natural generated phenomena such as hurricane, tornado and earthquake.

The below grade wall thickness was established by applying a uniform lateral pressure at any given depth, (i.e., sum of soil and water pressures). For the rock site, the soil pressure was assumed based on a unit weight of rock of 2720 kg/m3 and an at-rest lateral pressure coefficient Ko of 0.3. For the deep soil site, the soil pressure was assumed based on a unit weight of 2165 kg/m3 and a Ko of 0.5. The unit weight of water was assumed at 1000 kg/m3. Both lateral soil and water pressures were assumed to increase linearly with depth. The circular wall thickness was estimated by assuming the uniform pressure was resisted by the exterior wall acting as a compression ring. The straight wall was assumed to span vertically between five major floor levels, which were assumed to act as diaphragms. Under these assumptions, the thickness of the exterior wall increased with the depth of embedment. In general, the circular wall was thinner than the straight wall associated with the square structure. In the cost comparison, the cost of the circular wall was increased to account for the increased difficulty of placing forms and reinforcement in this wall system. Also, the cost of placing a cubic meter of concrete was increased with depth of placement.

The rock excavation was assumed to require blasting. The rock face was assumed to taper at 4 V to 1 H between benches (terrace levels). Benches 3 m wide were assumed at 9 m vertical intervals. Rock anchors were assumed to be required to keep the rock face stable. Also, a thin layer of shotcrete placed over welded wire mesh was assumed to be applied to the rock face to stabilize the rock surface.

The soil excavation was assumed to occur within the confines of a 1 m thick diaphragm wall of reinforced concrete. This wall was assumed to be constructed similar to a slurry wall placed to the required depth. As the excavation progresses, walers (horizontal beams) are placed on the inside face of the wall to resist soil and water pressures. The walers become larger and




more closely spaced as the depth increases. While the study provided a comparison at full embedment, it is unlikely that the fully-embedded reactor building will be stable due to the uplift pressure generated by water at this depth. In addition, constructing the diaphragm wall to a full embedment depth would at best very costly and difficult. Other methods of excavation support would need to be explored.

The cost analysis [69] looked at embedment depths from 9 m to 65.8 m in approximately 9 m intervals.

The relative results of this cost benefit analysis are shown in Table 21and Figure 33.Costs identified consist of major elements that would be expected to vary depending on the selected embedment depth. The major elements of cost include excavation, dewatering, and foundations, as well as above and below grade exterior concrete walls. The cost analysis shows that regardless of the site conditions, the cost of the embedded structure increases with depth of embedment.

Table 21 Cost Analysis(all in k $'s)

EmbeddedDepth (m)

CircularExterior Wall, Rock Below

Grade

CircularExterior Wall,

Soil Below Grade

SquareExterior Wall, Rock Below

Grade

SquareExterior Wall,

Soil Below Grade

9 $30,517 $22,033 $32,227 $24,191

18 $45,028 $31,009 $48,614 $36,178

27 $70,022 $45,787 $77,481 $57,625

36 $112,342 $68,202 $125,304 $87,761

45 $180,872 $100,685 $202,204 $133,368

54 $287,502 $146,137 $326,660 $197,189

65.8 $451,773 $213,090 $513,311 $286,741




ESTIMATED CAPITAL COST - KEY ELEMENTS

$0

$100,000

$200,000

$300,000

$400,000

$500,000

$600,000

0 10 20 30 40 50 60 70

Embedment Depth (M)

Dol

lars

(in

000'

s)

Circular Exterior Wall, Rock below Grade

Circular Exterior Wall, Soil below Grade

Square Exterior Wall, Rock below Grade

Square Exterior Wall, Soil below Grade

Figure 33 Estimated Capital Cost of Key ElementsThe relative first cost option is just one of many attributes to be considered in selecting

the depth of embedment. For example, as the amount of the safety related equipment located below grade increases, the reactor building becomes less susceptible to emerging threats. Tables26 and 33 show the attribute scoring criteria and the embedment depth ranking summary, respectively.

B1.6.2 Life Cycle Cost

The above discussion is directed towards key elements of the capital cost that might vary with the selection of embedment depth. It recognized the there may be variation in life cycle cost for operations as well. Section B1.1 indicates that a partial embedment might result in a slight reduction in travel time to reach locations in the building to perform inspections and maintenance. This is difficult to quantify at this point without more details on the layout of equipment in the building. It is recommended that operations and maintenance needs be revisited during conceptual design to verify that the partial embedment alternative is indeed favorable with respect to access and travel times.

It could be speculated that full or partial embedment might reduce HVAC system loads and associated energy costs with reduction of solar and transmission heat loads. HVAC system




sizing and performance requirements are not usually driven significantly by external conditions.Concrete structures have substantial thermal inertial to dampen effects of external temperatures.Also, HVAC flows are driven more by internal loads, fresh air, and contamination control needs.Again, it is recommended that this concern be reviewed in more detail during conceptual design.

B1.7 MALEVOLENT HAZARDS

Following the attack of September 11, 2001, the NRC issued a number of interim compensatory measures (ICMs) to better define the security requirements for operating nuclear plants. The bulk of these ICMs are Safeguards Information. The requirements in the ICMs cover a variety of security topics including plant physical design, access controls, guard force management and fitness for duty. The ICMs were developed to be consistent with the overall response to potential threats as coordinated by the United States Department of Homeland Security. This overall response strategy assigned different parts of the overall threat spectrum to different agencies such as the Department of Defense, the Department of Homeland Security, the Federal Bureau of Investigation, the NRC and others. The resultant DBT for nuclear power plants is discussed above in Subsection A1.3.7. To ensure a more comprehensive approach to security at nuclear power plants the NRC asked licensees to assess their response and mitigation design features and response strategies to threats and consequences beyond the DBT. Examples of these scenarios are large fires and explosions and the loss of all spent fuel cooling.

Orders have been issued to organizations with active Design Certification applications to assess their design’s resistance to the crash of a large commercial aircraft. The orders included aforce/time curve to be used as input for the assessment. Success is based upon the assessment showing that core cooling or containment integrity and spent fuel cooling or spent fuel containment integrity is maintained. For plants without a containment, a more elaborate justification of no significant radiation release is required. Assessment guidelines are included inthe “Methodology for Performing Aircraft Impact Assessments for new plant Designs soon to be issued as NEI 08-13 [26]

For the reactor building embedment, there is a trade off to be made for this assessment.If the reactor building is totally underground, it presents no target for aircraft crash and no further assessment is needed other than for crash induced vibration effects on in-plant equipment.If the reactor building has portions above grade, then a beyond design basis assessment must be made in accordance with Reference 9 to show no loss of core or spent fuel cooling orcontainment integrity. To date, plant designs with reactor buildings incorporating robust structural design for seismic and DBT resistance can generally show acceptable aircraft crash assessment results. Also in general, these designs must be reviewed to ensure that protection is provided from all directions.

In addition, event mitigation features and licensee strategies must be provided to NRC during the Combined License application phase of plant deployment in accordance with proposed Rule 10CFR50.54(hh)2 [8].




Large fires are defined as those greater than those established in a conventional designbasis fire hazards analysis. These large fires than develop with large quantities of fuel or flammable chemicals, can affect large portions of the plant including multiple fire zones, well in excess of those having their origin from on-site sources. No explicit initiation mechanism is assumed for these fires. Historically, one source of these large fires is the release of jet fuel from a large commercial aircraft crash. This has led to the need to include jet fuel as part of the large fire assessment. Assessments involve ensuring that the capability to shut down and cool down the plant and spent fuel facility is not lost nor is there significant radiation release as a result of this large beyond design basis fire. Success is based upon having readily available mitigation equipment.

No explicit mechanism is assumed for loss of all spent fuel cooling. The assumed starting point (for LWRs) is instantaneous loss of spent fuel pool water. This portion of the beyond design basis requirement should not apply to the reactor building since if houses no spent fuel in bulk. In addition, when the spent fuel storage facility is assessed for loss of cooling, new scenarios will be required since PBMR NGNP does not use water for cooling.

Rules of thumb and guidelines that, if followed, will make the reactor plant more resistant to beyond DBT are:

1. Minimize or eliminate the effective target area for aircraft crash2. In addition to general structural robustness of the reactor building for seismic, enhance

the penetration resistance of reactor building structures

3. Maximize the inherent and passive fuel cooling and radiation containment features of the plant

4. Maximize separation of redundant safety function initiation features

5. Maximize separation of safety and defense in depth system features

6. Ensure that alternate sources and means of fuel cooling are available following a beyond design basis threat event

B1.8 NATURAL PHENOMENON HAZARDS

B 1.8.1 Introduction

In general, it is proposed that the design of the NGNP Plant for natural phenomena eventsbe carried out in conformance with the EPRI – Advanced Light Water Reactor Utility Requirements Document Revision 8 dated March, 1999 (EPRI – URD)[24]. This document should be updated to reflect the latest codes and standards as almost ten years have passed since the last revision to this document.




In general, it is the intent that the NGNP Plant would be constructed in the United States in most areas east of the Rocky Mountains. The following discussion reflects this assumption.

B1.8.2 Tornado Hazards

The current EPRI-URD calls for a design wind load of 49 meters per second (m/s) adjusted to a 100 year mean recurrence interval through the use of an importance factor of 1.11.Today, it is likely that the building code used by the local authority will be the International Building Code. The International Building Code provides a significant amount of wind design data and criteria but states that wind loads shall be determined in accordance with Section 6 of ASCE 7 - Minimum Design Loads for Buildings and Other Structures [61]. Based on the current ASCE 7 standard, the 49 m/s basic design speed would preclude the plant from being built along many sections of the eastern and gulf coast of the United States. A design wind speed of 65 m/s (3 second gust) 10 m above grade with a design exposure C adjusted by a 1.15 factor for safety related structures and 1.0 for non-safety related structures appears more appropriate for the NGNP plant design.

Following is a comparison of the EPRI-URD design parameters for tornado design versus NRC Regulatory Guide 1.76 dated March 2007 [62]:

Parameter EPRI-URD RG-1.76Max Tornado Wind Speed 134 m/s 103 m/sMax Rotational Speed 107 m/s 82 m/sMax Translational Speed 27 m/s 21 m/sRadius of Max Rotational Speed 45.7 m 45.7 mMaximum Pressure Drop 138 mb 83 mbRate of Pressure Drop 83 mb/sc 37 mb/s

The EPRI-URD data is more conservative than RG 1.76 [62] in terms of pressure drop and wind velocity; therefore, it is proposed to use RG 1.76 [62] criteria. This criteria addresses tornado wind and pressure drop anywhere in the continental United States.

RG 1.76 [62] specifies the following tornado generated missiles be considered in the design of safety related nuclear structures:

Schedule 40 Pipe (0.168 m dia. x 4.58 m long) 130 kg 41 m/sAutomobile (5 m x 2 m x 1.3 m) 1810 kg 41 m/sSolid Steel Sphere ( 2.54 cm diameter) 0.0669 kg 8 m/s

The automobile is not considered in heights greater than 9.14 m above grade.




B1.8.2 Flooding Hazards

The site shall be chosen so that the flood level including the flood from a potential dam break be kept at a minimum 0.3 m below existing grade.

B1.9 NATURAL GEOLOGICAL HAZARDSNatural geologic hazards include seismic-related hazards, non-tectonic site deformation,

and volcanic hazards. A brief listing is given below based on references [46] and [47]

B1.9.1 Tectonic and Seismic Hazards

Seismic-related hazards include site earthquake ground shaking, tectonic site deformation (fault rupture and associated tectonic surface deformation, failure induced by high tectonic stresses), ground failure induced by ground shaking including liquefaction, differential compaction and land-sliding, and earthquake-induced flooding. For sites adjacent to large bodies of water, hazards include tsunami and seiche.

In general, it is proposed that seismic design be performed in accordance with EPRI-URD [24]. Highlights of the design requirements are discussed below as well as changes imposed by the latest design codes and or standards. It is expected that a seismic Basis of Design document will be prepared early in conceptual design to capture this methodology.

B1.9.1.1 Seismic ClassificationThe seismic classification system will be consistent with Nuclear Regulatory Guide 1.29

[70]. EPRI-URD [24] has added Seismic Category II to address non-seismic items/structures whose collapse could jeopardize the loss of function of Safety Class components or structures.Seismic Category II requires that only structural integrity be maintained, not functional integrity.Consequently, the seismic classifications are as follows:

• Seismic Category I – This classification includes all structures, systems and components whose safety class is SC-1, SC-2, or SC-3. Seismic category I shall also include spent fuel storage pool structures including all fuel racks.

• Seismic Category II – This classification applies to all plant structures, systems and components which perform no nuclear safety function but whose failure could degrade SC-1, SC-2, and SC-3 structures, systems and/or components.

• Non-seismic Category III– This classification includes all structures that do not fall into Seismic Category I or Seismic Category II structures.




B1.9.1.2 OBE Design Basis

EPRI-URD [24] has eliminated the operating design earthquake from its load cases. The Nuclear Regulatory Commission (NRC) has essentially agreed to this change (NRC Policy Issue I.M., “Elimination of OBE); however, there continues to be discussions between EPRI and the NRC concerning the number of one-half Safe Shut-Down Earthquake (SSE) cycles to be used in the evaluation of equipment and components in regard to fatigue and seismic performance.

B1.9.1.3 Ground Motion Characteristics

EPRI proposes to use a SSE comprising a single ground motion spectrum conforming to Regulatory Guide 1.60 [63] anchored to 0.3g peak ground acceleration and applied at the free-field soil surface except at site where rock extends above the nuclear island founding level where the peak ground acceleration is applied at the top of rock. This peak ground acceleration will allow the NGNP to be constructed at most sites in the continental United States east of the RockyMountains.

The design response spectra shall be in accordance with Regulatory Guide 1.60 –“Design Response Spectra for Seismic Design of Nuclear Power Plants” [63] with a time history to envelop the design spectra.

The project will initially use a deterministic seismic approach that will provide an enveloping ARS for a standard NGNP design. Specific site criteria, when available, will be checked against this enveloping ARS.

Once a site is selected, the vibratory ground motion characteristics will be evaluated in accordance with the requirements set forth in Section 100.23, "Geologic and Seismic Siting Criteria," of Title 10, Part 100 [72], of the Code of Federal Regulations (10 CFR 100), "ReactorSite Criteria." [57].[54] Regulatory Guide 1.165 “Identification and Characterization of Seismic Sources and Determination of Safe Shutdown Earthquake Ground Motion” [71] provides Seismic Hazard Analysis (PSHA) for sites in different parts of the U.S. An acceptable method for determining the annual probability of exceeding the SSE, defined as the reference probability, is described in Appendix B of Regulatory Guide 1.165 [71]. The development of seismic hazard results will be based on a site-specific PSHA and will consider site amplification effects.

The NRC also provides an alternative approach to satisfy 10 CFR 100.23 [72] and Appendix S, “Earthquake Engineering Criteria for Nuclear Power Plants” to 10 CFR Part 50 [76]. This method, detailed in Regulatory Guide 1.208 “A Performance-based Approach to Define the Site-Specific Earthquake Ground Motion” [73], provides guidance for development of the site specific ground motion response spectrum (GMRS). The performance based approach combines ground motion hazard with equipment/structure response to establish risk-consistent GMRS. It




differs from the hazard-consistent ground shaking that would be determined from the hazard reference probability described in Regulatory Guide 1.165 [71].

The methodology for developing the GMRS is consistent with ASCE/SEI Standard 43-05, “Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities”[74]. The method is based on the use of site-specific mean seismic hazard. The ASCE/SEIStandard 43-05 graded approach is designed to meet quantitative safety performance goals for a range of nuclear facilities.

B1.9.1.4 Analysis & Design

Design of Seismic Category I structures shall follow the ground motion characteristics specified above, the analytical techniques specified in ASCE Standard 4-98 [41] and meet all quality assurance requirements of 10 CFR 50 [54], Appendix B. ASCE Standard 4-98 has been accepted by the NRC as meeting the requirements for seismic analysis with some exceptions.Design of Seismic Category I concrete structures shall be in accordance with ACI 349-06 [64].Design of Seismic Category I steel structures shall be in accordance with AISC N690L-03 [65].This version of the AISC code for safety related steel structures for nuclear facilities is a LRFD version of the code which is in line with the concrete design code for nuclear facilities.

Design of seismic category II structures shall utilize the same analysis and design codes as those used for seismic category I structures; however, these structures may be constructed to the requirements of the non-nuclear building codes. In general, these requirements will invoke the International Building Code (IBC-06) [66] or the version specified by the local building authority.

Design of Seismic Category III structures shall follow the analytical requirements of IBC-06 [66]. IBC-06 specifies that ACI-318-05 [67] by used for the design and construction of concrete structures and that AISC – ASD/LRFD Steel Construction manual, 13th Edition [77] beused for steel design and construction. As an alternate, the use of ASCE Standard 7 for seismic analysis will be explored.

ASCE 43-05 [74] is a seismic design criteria for structures, systems and components in nuclear facilitates and is intended for use in conjunction with the design and analysis references listed above. ASCE 43-05 is similar to DOE-STD-1020-2002 [75]. ASCE 43-05 has been accepted by RG 1.208 [73] (DG 1146) [78] for defining site specific design-basis earthquake response spectrum. ASCE 43-05 [74]may be used to set different levels of seismic input for structures, systems and components that have different failure consequences.




B1.9.1.5 Soil Structure Interaction

An additional consideration of this study is to evaluate the effects of structure embedment on the seismic input ground motion at the base of the structure. In general, embedding the structure reduces the input motion at the base of the structure; thereby, reducing the seismic load.However, there are limits to this effect. Based on published literature references [38], [40] and[38][41], it is reasonably conservative to assume that embedment will not reduce the input base ground motion to less than approximately 60% of the surface ground motion. For the assumedreactor building, this limit is reached at a depth of approximately 30 m. In addition, ASCE 4 [41] states that the top 6 m or one-half of the embedment depth, whichever is less, must be neglected in the soil structure interaction formulation. Based on the above, it is estimated that little or no reduction will be realized at embedment exceeding 36 m or about 50% embedment of the reactor building. It is not unrealistic to assume that the input motion at the base of the structure varies linearly from 100% at grade to 60% at 36 m.

An example of this beneficial effect is illustrated in Figure 34 and Table 22 below.

There are many factors that contribute to amplification of the peak ground acceleration.The building stiffness, aspect ratio, and layer thickness of the soil along with earthquake characteristics significantly impact the amplification factor.

Figure 34 Reactor Building Seismic Accelerations




Table 22 Hypothetical Reactor Building Seismic Accelerations

REACTOR BUILDING EMBEDMENT

PERCENTANGE

ESTIMATED INPUT ACCELERATION AT BASE OF

REACTOR BUILDING (g’s)

ESTIMATEDPEAK

ACCELERATIONTOP OF

REACTORBUILDING

0 0.300 1.0 (ASSUMED)

10 0.276 0.92

20 0.252 0.84

30 0.228 0.76

40 0.204 0.68

50 0.180 0.60

100 0.180 0.60

In the above table for each embedment depth, column 2 represents the adjusted baseacceleration from a peak ground level value of 0.3g. Column 3 is based on an assumed value of 1.0 corresponding to 0.3 g peak ground level acceleration. This value would ultimately bedetermined based on site specific seismic conditions and characteristics of the structure. The amplified acceleration at the top of the building varies proportionately with the acceleration at the base.

B1.9.2 Non-Tectonic Surface Deformation

Non-tectonic phenomena that can relate to surface deformation at a site include glacially-induced faulting, growth faulting, landslides or other mass-wasting phenomena, collapse or subsidence in areas due to underground voids such as found in karstic limestone terrain, subsidence due to excessive withdrawal of fluid (such as oil and gas), chemical weathering, and induced seismicity and fault movement caused by reservoir impoundment and fluid injection and removal.

B1.9.3 Volcanic Hazards

Potential volcanic hazards may include: lava flows, ballistic projections, ash falls, pyroclastic flows and debris avalanches, lahars (mudflow composed of pyroclastic material and water that flows down from a volcano, typically along a river valley) and flooding, seismic activity, ground deformation, atmospheric affects, and acid rains and gases. These phenomena are restricted to limited areas in the western United States.




B1.9.4 Effects on Embedment Study

The hazards listed above have to be evaluated and addressed in the licensing documents for a nuclear facility. For the purpose of this study, however, only those phenomena that differently affect embedded and non-embedded structures need be considered. These phenomena are seismic hazard and liquefaction, both of which are reduced by partiallyembedding the structure.

As discussed in Section B1.6, embedding a structure decreases the input motion at the base of the structure, thereby reducing the dynamic forces on the structure. Certain limits apply to this benefit. For example, NRC Standard Review Plan 3.7.2 requires the horizontal component of the acceleration at foundation depth to be no less than 60 percent of that at finished grade in the free field.

Liquefaction refers to generation of high excess pore water pressure as a loose, saturated granular soil is forced to assume a denser configuration. The excess pore water pressure reduces the shear strength of the soil. This can lead to loss of foundation support, or a flow slide on sloping ground. Embedment reduces the potential for seismically-induced liquefaction. One reason is that the ground motion decreases with depth below the ground surface, reducing the energy available to shake the loose soil into a denser configuration. Below depths of 15 to 20 m, liquefaction is generally not an issue for moderate earthquakes. If liquefaction were an issue, the soil could be densified using standard techniques, prior to building the structure.

B1.10 CHEMICAL RELEASE, EXPLOSIONS, MANMADE HAZARDS

Consideration needs to be given to the hazards of chemicals stored and or transported in range of the Reactor, the potential effects, and mitigating features that could be utilized. This type of concern does often materialize with operating facilities located near active industrial areas.

B1.10.1 Toxic Chemical Releases

Appendix A, “General Design Criteria for Nuclear Power Plants”, of 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities”[54] requires that the design of structures, systems, and components (SSCs) important to safety be able to accommodate the effects of and be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents. Criterion 19 of Appendix A requires that a control room from which actions can be taken to operate the facility safely under normal conditions and to maintain it in a safe condition during accidents. Releases of onsite and offsite hazardous chemicals, which infiltrate the control room, can result in the control room becoming uninhabitable.




Regulatory Guide 1.78, “Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release” [35], describes assumptions acceptable to the NRC staff for use in assessing the habitability of the control room during and after a postulated external release of hazardous chemicals from mobile or stationary sources, offsite or onsite, including storage tanks, pipelines, tank trucks, railroad cars, and barges. The regulatory guide directs that all hazardous chemicals present onsite in weights greater than 100 pounds within 0.3 mile of the control room or chemicals present within a 5-mile radius of the plant above certain threshold quantities be considered in a control room evaluation. The consideration of mobile sources within 5 miles of the plant is also dependent on the shipment frequency. The evaluation is performed to determine the potential toxic effects on control room operators from accidental chemical releases reaching the control room fresh air intake.

The regulatory guidance states that two types of chemical accidents should be considered for each source of hazardous chemicals: maximum concentration accidents; and maximum concentration-duration accidents. A maximum concentration accident is one that results in a short-term puff or instantaneous release of a large quantity of hazardous chemicals. The typical assumption is the failure of the largest single container resulting in the instantaneous release of the total contents of the tank or vessel. Depending on the boiling point of the chemical, thevapor release may be the result of just puddle evaporation or vaporization or may involve an instantaneous 3-dimensional gaseous puff release due to flashing followed by puddle vaporization. A maximum concentration-duration accident is one that results in a long-term,low-leakage-rate release. For a maximum concentration-duration accident, the continuous release of hazardous chemicals from the largest safety relief valve on a stationary, mobile, or onsite source should be considered.

According to Regulatory Guide 1.78 [35], the atmospheric transport of a released hazardous chemical should be calculated using a dispersion or diffusion model that permits temporal as well as spatial variations in release terms and concentrations. Atmospheric dispersion models can be used for dispersion calculations as long as these models are capable of calculating spatial and temporal variations in release terms and concentrations, simulating building wake effects, and simulating near-field effects. The NRC uses a computer code, HABIT, for control room habitability evaluation. The HABIT code is described in NUREG/CR-6210, "Computer Codes for Evaluation of Control Room Habitability (HABIT)". This code has two modules, EXTRAN and CHEM, for calculation of chemical concentration and exposure, respectively. The model in EXTRAN, a Gaussian plume or puff dispersion model, allows longitudinal, lateral, and vertical dispersions. The model also allows for the effect of wakes and for additional dispersion in the vertical direction when the distance between the release point and the control room is small. Other atmospheric dispersion models (e.g., ARCON96, Ramsdell and Simonen, 1987) with similar capabilities may be used for dispersion calculations.

The atmospheric dispersion model provides a time history in the control room of the accidentally released toxic chemical concentration based on the assumed meteorological conditions and control room design parameters (i.e., control room free volume, normal andemergency ventilation rates, and time to isolate the control room). The meteorological condition




assumed in such an analysis is the 5-percentile worst case dispersion condition at the site to provide a conservative assessment. The calculated chemical concentrations in the control room are then compared to health effects data (i.e., toxicity limit) appropriate to determine potential control room operator incapacitation. Regulatory Guide 1.78 recommends the use of the Immediately Dangerous to Life or Health (IDLH) concentration provided by the National Institute for Occupational Safety and Health (NIOSH, 1997)[30]. The IDLH is an atmospheric concentration of any toxic, corrosive or asphyxiate substance that poses an immediate threat to life or would cause irreversible or delayed adverse health effects or would interfere with an individual's ability to escape from a dangerous atmosphere based on a 30-minute exposure.

According to NUREG/CR-6944, “Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs)”, Volume 6 [34], certain processes for the production of nuclear hydrogen may have large inventories of toxic chemicals. One example is the sulfur iodine process that uses the toxic chemicals hydrogen iodine (HI) and iodine (I2). Under the appropriate accident conditions, a toxic gas release could be postulated. In addition, many corrosive chemicals are toxic but, in most cases, their corrosive characteristics dominate the hazard. Corrosive chemicals are a hazard to both equipment and operators whereas toxic chemicals are primarily a hazard to the control room operator. Ultimately, the impact on the control room operators will be dependent on the quantity of chemical released, the release phenomenology, meteorological conditions, and the toxicity of the chemical.

One method of mitigating potential adverse effects of accidental toxic chemical releases on control room operators is the use of instrumentation to detect the presence of the chemical at the control room air intake whereby the operators can take protective actions such as donning breathing apparatus or the control room may be automatically isolated upon detection. A protective design feature may include dual control room air intakes that are sufficiently separated so as to prevent a toxic plume from affecting both intakes simultaneously.

B1.10.2 Asphyxiant Gas Releases

Another potential source of concern is the use of gases at industrial facilities that are generally considered to be non-toxic but that could adversely affect human health and equipment operability by displacing sufficient atmospheric oxygen. One example includes pressurized helium fluid in the intermediate heat transfer loop whereby a helium leak in a confined space could create an environment incapable of sustaining life and in which personnel would be incapacitated within seconds. Other potential asphyxiants that could be found in a chemical process includes nitrogen and carbon dioxide. Displacing air with an inert gas also has the capability of incapacitating equipment. For example, if emergency diesel generators are needed for backup power in the event of a loss of offsite power, they may not be able to operate due to an inert gas plume at the air intake reducing oxygen content. Potential mitigation measures could include locating the inert gases far from the nuclear plant and keeping local inventories of these gases at a minimum.




The release phenomenology of the inert gas determines how far from the source of the leak asphyxiants can have an adverse effect on operators and equipment. If the gas or cryogenic liquid release has a vapor density greater than that of air, the heavier-than-air cloud can accumulate at ground level and travel well away from the source. Locating the inert-gas storage well away from the plant and minimizing vessel inventory are effective ways to reduce the risk of adverse effects of asphyxiating plumes on the nuclear plant personnel and equipment.

B1.10.3 Flammable Releases

According to NUREG/CR-6944 [34], most hydrogen is produced by steam reforming of natural gas. As this is an endothermic process for hydrogen production, about 30% of the natural gas is used to produce high-temperature heat to drive the chemical reactions. Because the process involves large quantities of natural gas, some nuclear hydrogen processes have the potential for large-scale release of flammable gases. In addition, high-temperatures from nuclear reactors are being considered as a heat source for oil refineries, shale oil and tar sands production facilities, and coal gasification and liquefaction processes. All of these involve the production of flammable gases and/or liquids in large quantities. Flammable fluids will be present in process streams at elevated pressures and at high (or low) temperatures. There will also inevitably be on-site storage of intermediates and products, possibly in large tanks.

The potential impacts on the safety-related reactor plant SSCs may involve a leak of flammable fluid resulting in an explosion with blast effects on the nuclear plant or a flammable fluid leak resulting in a fire with significant heat flux that could damage the nuclear plant.Another potential impact is operator injury or impairment from burns due to high heat fluxes.

Heat flux is generally the most serious hazard posed by a flammable release. While hydrogen flames give off relatively little thermal radiation, hydrocarbon flames have intense heat fluxes due to their high heat content. The accidental release of a flammable gas or vapor could also result in an unconfined vapor cloud explosion (UVCE). If the accidental release involves a liquid above its atmospheric boiling point, the result could be a boiling liquid expanding vapor explosion (BLEVE). Blast effects from an UVCE or BLEVE in the coupled chemical process plant could conceivably impact the nuclear plant through incident overpressure or debris missiles. The blast could occur either at the site of the released flammable chemical or at some point downwind as the chemical disperses to within the flammable concentration range given an ignition source (i.e., delayed ignition).

The NRC provides guidance on the evaluation of the impact of explosions in Regulatory Guide 1.91, “Evaluations of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants” [36]. The method involves the determination of the distance from the potential release point to the plant beyond which the blast overpressure from an explosion is not likely to have an adverse impact. This distance is based on a peak positive incident overpressure of 1 pound per square inch (psi) below which no significant damage to SSCs of concern is expected by the NRC. The distance is a function of the 1/3rd power of the TNT-equivalentweight of the released substance.




The blast overpressure impacts, as well as the heat flux effects, can also be evaluated using the U. S. Environmental Protection Agency (EPA) computer model ALOHA (Areal Locations of Hazardous Atmospheres) [33]. ALOHA allows the user to estimate the concentrations due to downwind dispersion of a chemical cloud based on the physical characteristics of the released chemical, atmospheric conditions, and specific circumstances of the release. ALOHA was jointly produced by the National Oceanic and Atmospheric Administration (NOAA) Hazardous Materials Response and Assessment Division and the EPA Chemical Emergency Preparedness and Prevention Office (CEPPO). It was originally based on a simple model with a continuous point source with a Gaussian plume distribution but now also includes a heavy gas dispersion model. The heavy gas dispersion model used by ALOHA is theDense Gas Dispersion (DEGADIS) model that was originally developed for the U.S. Coast Guard and the Gas Research Institute (GRI), primarily for simulation of the dispersion of cryogenic flammable gases.

As in the case of asphyxiate gas releases, flammable hazards can be can be mitigated by locating the flammable liquids far from the nuclear plant and by keeping local inventory at a minimum. According to a report prepared by Idaho National Laboratory [32] on separation requirements of a hydrogen production plant and high-temperature nuclear reactor, storage of 100 kg or less of hydrogen should kept at least 110 meters away from the nuclear plant without mitigation and 60 to 120 meters away if a blast barrier is constructed between the two facilities.The report also recommends that the major nuclear plant structures be placed below ground level as an effective mitigation measure.

Consideration should also be given to the fact that heat flux from hydrocarbon fires is more intense than from a .hydrogen flame and the flammability range (i.e., upper and lower flammability concentrations) is smaller for hydrocarbons than it is for hydrogen.

Again, the release phenomenology of the flammable gas or liquid determines how far from the source of the leak an explosion or fire can occur. When the gas density is greater than that of air, flammable concentrations can accumulate near ground level and travel away from the source. Plumes of flammable vapors can travel a considerable distance but the use of containment walls or other impediments to flow can reduce the risk. Locating storage tanks as far as possible from the plant and minimizing tank inventory are also effective ways to diminish the risk of plumes that could affect the nuclear plant.

B1.10.4 Oxygen Releases

As discussed in NUREG/CR-6944 [34], a nuclear hydrogen plant converts water into hydrogen and oxygen, except for nuclear heat for steam reforming of natural gas, with oxygen as the byproduct. The oxygen may be sold if there is a local market or it may be vented to the atmosphere. Oxygen that is sold may be transported to storage sites or to the customer by




pipeline. In such cases, the site inventory will be limited to the inventories within the hydrogen process.

The potential effects of an accidental oxygen release include damage, wear, or impairment of the safety-related reactor plant SSCs due to the contact of oxygen enriched air with combustible materials leading to a fire or degrading equipment over time. Another potential impact is operator injury from burns caused by the increased flammability of combustible materials in the oxygen enriched environment.

Nuclear hydrogen plants will produce large quantities of pure oxygen that escapes fromthe chemical plant and cools as it is depressurized, potentially creating a heavy gas cloud that can flow at ground level to the reactor. In addition, spontaneous combustion becomes likely for many materials with pure oxygen and materials such as steel that are normally considered as noncombustible can burn in an oxygen rich atmosphere.

Another concern is related to the plant continuously releasing oxygen, leading to locally higher oxygen concentrations that may have secondary impacts. The oxygen may be stored in very large quantities for some applications. The important considerations in terms of nuclear plant safety are the storage quantities, temperatures, and pressures of the oxygen.

B1.10.5 SummaryThe discussion of the application of Regulatory Guide 1.78 [35] requirements regarding

the impact of postulated toxic chemical releases on control room habitability may not apply to the NGNP. Chapter 10 of the NGNP PCDR indicates that control room operators are not required to respond to design basis events and their actions are not credited in safety analyses. In this case, Criterion 19 of Appendix A of 10 CFR Part 50 [54] would not be applicable to the NGNP, precluding consideration of the requirements of Regulatory Guide 1.78 [35]. The consideration of the effects of accidental toxic chemical releases on the NGNP would be then be confined to adverse health effects on plant personnel due to the propagation of hazardouschemicals through secondary or tertiary systems to the reactor building or on plant equipment in the case of asphyxiates such as emergency diesel generators.

In regard to flammable/explosive chemicals, NRC Regulatory Guide 1.91, “Evaluations of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants” [36]would apply to the NGNP. The guide provides a method acceptable to the NRC staff that involves the determination of the distance from the potential flammable/explosive chemical release point to the plant beyond which the blast overpressure from an explosion is not likely to have an adverse impact. This distance is based on a peak positive incident overpressure of 1 pound per square inch (psi) below which no significant damage to SSCs of concern is expected by the NRC. If the distance is smaller than that determined by the Regulatory Guide, further analysis of the effects of the flammable/explosive chemical would need to be conducted, possibly including a PRA.




Both toxic and flammable/explosive hazards can be mitigated by locating the hazardous chemicals far from the nuclear plant and by keeping local inventory at a minimum, as well as embedding the major nuclear plant structures below ground level. However, a potential drawback of embedment is that the NGNP could be more vulnerable to the intrusion of toxic and/or flammable heavier-than-air gases.

B2 EVALUATION AND RECOMMENDATIONREGARDING REACTOR EMBEDMENT

B.2.1 OBJECTIVE RANKING

Table 23 below characterizes the various objective considerations discussed previously.Weights are assigned defining the relative importance of each objective. Items with higher weights are considered major drivers in the decision process. Items with lower weight are considered less significant as a discriminator in making the selection of embedment depth.

Table 23 Objective Ranking and Relative Weights

Objective WeightDesign for Operations and Proper System Mechanical Functions

Operational Needs (Safety equipment layout) (Note 1) 10Safety, Investment Protection, and Security

Heat Dissipation to Environment (Note 2) 5Design Basis Threat and Malevolent Hazards (Note 3) 15Natural Phenomena (Note 4) 15Natural Geological Phenomena (Note 5) 10Chemical Releases and Explosions (Note 6) 5

Cost and Construction ComplexityCost Benefit Consideration (Note 7) 10Water Table Effects (Note 8) 10Geotechnical Constraints and Foundation Performance (Note 9)

10

Construction Considerations (Note 10) 10

Total Weight Percent 40

Notes: 1) Design for Operational effects (Reactor protection, access for refueling, etc.) for normal and emergency operations, and maintenance activities. Relative locationsof sub-systems with respect to reactor are considered. This is an important consideration. However, this weight is moderate because there is very little difference in the impact of operations on the selection of embedment.

2) The ability of the building to transfer heat to the environment either to ground or to ambient air will be assessed. The reason this weight assigned to this item is




low is that heat transfer through walls to the environment is not a design basis event, also it is shown the there is very little numerical difference expected between heat transfer to soil or to air. This is primarily due to the thermal inertia of the concrete citadel walls.

3) The element addresses protection against the design basis safeguards and securitythreat DBT and the beyond design threat imposed by a potential non accidental aircraft strike, or other BDBE. This item carries a high weight due to the importance of safeguards and security in the licensability of the plant.

4) Natural Phenomenon Hazards (NPH), such as tornado (including missiles) and flooding will are considered. This carries a high weight because its impact on safety and licensibility of the plant. It is common practice to design the exterior walls of a nuclear plant to accommodate these hazards.

5) The potential impacts that natural geological hazards such as dissolution features in soluble rock, weak compressible soils, slope instability resulting in landslide potential, and liquefaction and other earthquake induced flow phenomena are considered. This carries and moderate weight. The design challenges are not in surmountable regardless of embedment selection. Seismic response will be improved by a moderate amount by partial embedment.

6) Consideration for the hazards of chemicals stored and or transported in range of the Reactor, the potential effects, and mitigating features that could be utilized.This type of concern does often materialize with operating facilities located of active industrial areas. The potential for hydrogen explosions is addressed here.It is expected that robust structure required by other elements will facilitate design to accommodate this hazard. There are many options to control this hazard.

7) Cost benefit including key elements of relative capital cost will be developed for various embedment options and compared in a cost benefit table. This is a major element to consider and there is a very significant impact on cost associated with full embedment of the structure.

8) Groundwater levels are evaluated with regard to impact on structural design, construction dewatering schemes and need for waterproofing. This has a moderate impact on cost.

9) A range of foundation materials are considered with regard to static and dynamic properties and foundation performance. This has a moderate impact on cost.

10) The effects of the embedment depth on constructability using modern techniques such as modularization are considered. The excavation method used will require an evaluation of overburden thickness, elevation of competent rock, site space




constraints, and cost and schedule impacts. This element has a major impact on cost.

B2.2 LIST OF ALTERNATIVE SITES CONSIDEREDThe following types of sites are typically found throughout the U.S. and will be discussed

in this study. Rock sites typically have deep water tables. Soils sites with a deep water table exhibit similar characteristics to rock sites with respect to effect- or lack thereof of water. So the two selected site types are:

• Rock sites such as INL with a deep water table

• Deep soil sites with high water table

B2.3 ALTERNATIVE REACTOR EMBEDMENT CONCEPTSThe following alternatives are considered in this study for scoring and ranking purposes.

It is expected that the optimal embedment depth will be developed in conceptual design and will be dependent on specific site conditions and ongoing analysis of operations needs.

• Minimal embedment of the building at approximately 7 to 10 meters

• Partial embedment 20 to 30 meters

• Full Embedment 60+ meters

B2.4 SCORING CRITERIAThe following range of scoring is used to establish relative ranking of the alternatives to

satisfy objectives as shown in Tables 24 through 29 below.

Score of 0 Alternative does not satisfy objective

Score of 10 Alternative is clearly best suited to satisfy objective




Table 24 Alternative Ranking Rock Sites (Deep Water Table)Minimum Embedment

Objective Weight AlternativeSatisfaction

Score

WeightedScore

Design for Operations and Proper System Mechanical Functions

Operational Needs (Safety equipment layout) (Note 1)

10 6 60

Safety, Investment Protection, and Security

Heat Dissipation to Environment (Note 2)

5 6 30

Design Basis Threat and Malevolent Hazards (Note 3)

15 4 60

Natural Phenomena(Note 4)

15 4 60

Natural Geological Phenomena (Note 5)

10 4 40

Chemical Releases and Explosions (Note 6)

5 4 20

Cost and Construction ComplexityCost Benefit Consideration (Note 7)

10 8 80

Water Table Effects (Note 8)

10 5 50

GeotechnicalConstraints and FoundationPerformance (Note 9)

10 6 60

ConstructionConsiderations (Note 10)

10 6 60

Total Weight Percent 40 250




Table 25 Alternative Ranking Rock Sites (Deep Water Table)Partial Embedment


Score

WeightedScore



10 8 80



5 6 30


15 6 90

Natural Phenomena (Note 4)

15 6 90


10 6 60


5 6 30


10 6 60


10 5 50


10 5 50


10 8 80





Table 26 Alternative Ranking Rock Sites (Deep Water Table)Full Embedment


Score

WeightedScore



10 5 50



5 6 30


15 7 105


15 6 90


10 3 30


5 4 20


10 2 20


10 5 50


10 4 40


10 2 20





Table 27 Alternative Ranking Soil Sites (High Water Table)Minimal Embedment


Score

WeightedScore



10 6 60



5 6 30


15 4 60


15 4 60


10 4 40


5 4 20


10 8 80


10 10 100


10 8 80


10 6 60





Table 28 Alternative Ranking Soil Sites (High Water Table)Partial Embedment


Score

WeightedScore


Operational Needs(Safety equipment layout) (Note 1)

10 8 80



5 6 30


15 6 90


15 6 90

Natural GeologicalPhenomena (Note 5)

10 6 60


5 6 30


10 6 60


10 7 70


10 6 60


10 8 80





Table 29 Alternative Ranking Soil Sites (High Water Table)Full Embedment


Score

WeightedScore



10 5 50



5 6 30


15 7 105


15 6 90


10 3 30


5 4 20


10 2 20


10 0 0


10 2 20


10 2 20





B2.5 ALTERNATIVE RANKING SUMMARY

The following Table 30 provides a summary of scores for the design options from Table 24 through 29.

Table 30 Design Alternatives Ranking Design Alternative Total Weighted Score Remarks

Rock Site Minimum Embedment 510 Maximum exposure to hazards Rock Site Partial Embedment 610 Best Score for rock site with

balance of protection from hazards and reduced cost

Rock Site Full Embedment 465 Water table is not a concern.Excavation and foundation complexity contributes to

construction costSoil Site Minimum Embedment 580 Maximum exposure to hazards

but Benefit with reduced water table concern

Soil Site Partial Embedment 650 Best score for soil site Soil Site Full Embedment 395 Water table and foundation

complexity contribute to very high costs




B2.6 ROLE OF EMBEDMENT TO SATISFY T&FRS

Part A of this study defines reactor building functions and requirements that must be fulfilled independently of the extent of embedment. It also provides a preliminary set of licensing basis events involving leaks and breaks in the PHTS and SHTS piping that the building must withstand. It evaluates a number of alternative design strategies for mitigating pipe breaks and minimizing radiological releases from the building. These options can all be applied to any level of reactor embedment. Hence the functions and requirements section is not dependent on the outcome or conclusions of the reactor embedment section.

The reactor embedment section is somewhat dependent on the results of functions and requirements section. The reactor embedment section also describes external hazards and physical security requirements that the building design must be able to accommodate. The capabilities of the reactor to protect against these hazards can be significantly influenced by the level of embedment.

A complete correlation between identified T&FRs and the consideration of reactor embedment is included in Table 31. This table shows the full list of T&FR, the role, if any, that embedment serves to accommodate these requirements, and the rational for selecting a particular embedment alternative. T&FRs also shown in Table 6 in the functions and requirements section.

NG

NP-

NH

S 10

0-R

XB

LDG

Rev

0N

GN

P C

once

ptua

l Des

ign

Stud

y Rea

ctor

Bui

ldin

g Fu

nctio

nal a

nd T

echn

ical

R

equi

rem

ents

and

Eva

luat

ion

of R

eact

or E

mbe

dmen

t

NG

NP-

NH

S10

0-RX

BLD

G0

Fina

l09

1608

.doc

Sept

embe

r 15,

200

816

8 of

186

Tabl

e31

Rol

e of

Em

bedm

ent i

n Sa

tisfy

ing

Rea

ctor

Bui

ldin

g Fu

nctio

ns a

nd R

equi

rem

ents

T&FR

Num

ber

Req

uire

men

t R

ole

of R

eact

or E

mbe

dmen

t

NH

SB-1

.0Th

e sc

ope

of fu

nctio

ns a

nd re

quire

men

ts fo

r the

reac

tor

build

ing

shal

l be

allo

cate

d to

the

Nuc

lear

Hea

t Sup

ply

Build

ing

(NH

SB),

NH

SS H

VAC

Sys

tem

, and

NH

SB

Pres

sure

Rel

ief S

yste

m (P

RS)

as

desc

ribed

in th

e PC

DR

.

Not

App

licab

le th

is is

a G

ener

al R

equi

rem

ent

NH

SB-2

.0Th

e N

ucle

ar H

eat S

uppl

y Bu

ildin

g (N

HSB

) sha

ll pe

rform

th

e fo

llow

ing

func

tions

:

N

HSB

-2.1

Hou

se a

nd p

rovi

de c

ivil

stru

ctur

al s

uppo

rt fo

r the

reac

tor

and

all t

he S

SCs

with

in th

e sc

ope

of th

e N

HSS

(bu

ildin

g ge

omet

ry/s

pace

)

Parti

al e

mbe

dmen

t is

favo

red

with

redu

ced

seis

mic

resp

onse

for u

p to

50

perc

ent

embe

dmen

t. P

artia

l em

bedm

ent a

lso

favo

rabl

e fo

r ope

ratio

n an

d eq

uipm

ent l

ocat

ion.

See

se

ctio

n B1

.1 a

nd B

1.9.

NH

SB-2

.1.1

.Pr

ovid

e pe

rson

nel a

nd e

quip

men

t acc

ess

for p

lant

co

nstru

ctio

n, m

aint

enan

ce, o

pera

tion,

and

insp

ectio

n of

SS

Cs

in th

e N

HSS

and

for r

outin

e an

d em

erge

ncy

ingr

ess

and

egre

ss o

f pla

nt w

orke

rs

Parti

al e

mbe

dmen

t slig

htly

favo

rabl

e fo

r op

erat

ion

and

mai

nten

ance

as

wel

l as

for e

ase

of

cons

truct

ion.

See

sec

tion

B1.1

and

B1.

5.

NH

SB-2

.1.2

.Pr

ovid

e ra

diat

ion

shie

ldin

g fo

r pla

nt w

orke

rs a

nd th

e pu

blic

dur

ing

norm

al o

pera

tion

to k

eep

radi

atio

n ex

posu

res

ALAR

A

Not

affe

cted

by

embe

dmen

t.

NH

SB

-2.1

.3.

Lim

it ai

r flo

w in

neu

tron

field

s to

kee

p ro

utin

e re

leas

es o

f ac

tivat

ion

prod

ucts

and

oth

er re

leas

es d

urin

g no

rmal

pl

ant o

pera

tion

from

all

sour

ces

of ra

dioa

ctiv

ity in

side

the

NH

SB A

LAR

A

Not

affe

cted

by

embe

dmen

t.

NG

NP-

NH

S 10

0-R

XB

LDG

Rev

0N

GN

P C

once

ptua

l Des

ign

Stud

y Rea

ctor

Bui

ldin

g Fu

nctio

nal a

nd T

echn

ical

R

equi

rem

ents

and

Eva

luat

ion

of R

eact

or E

mbe

dmen

t

NG

NP-

NH

S10

0-RX

BLD

G0

Fina

l09

1608

.doc

Sept

embe

r 15,

200

816

9 of

186

T&FR

Num

ber

Req

uire

men

t R

ole

of R

eact

or E

mbe

dmen

t

NH

SB

-2.1

.4.

Mai

ntai

n in

tern

al e

nviro

nmen

tal c

ondi

tions

(tem

pera

ture

, hu

mid

ity, a

ir re

fresh

) for

NH

SS S

SCs

and

oper

ator

s.It

coul

d be

spe

cula

ted

that

em

bedm

ent m

ight

re

duce

HVA

C s

yste

m lo

ads

with

redu

ctio

n of

so

lar a

nd tr

ansm

issi

on h

eat l

oads

. H

VAC

sy

stem

siz

ing

and

perfo

rman

ce re

quire

men

ts a

re

not u

sual

ly d

riven

sig

nific

antly

by

exte

rnal

co

nditi

ons.

Con

cret

e st

ruct

ures

hav

e su

bsta

ntia

l th

erm

al in

ertia

l to

dam

pen

effe

cts

of e

xter

nal

tem

pera

ture

s. H

VAC

flow

s ar

e dr

iven

mor

e by

in

tern

al lo

ads

and

cont

amin

atio

n co

ntro

l nee

ds.

NH

SB-2

.1.5

.Su

ppor

t zon

ing

requ

irem

ents

for H

VAC

, rad

iatio

n, fi

re

prot

ectio

n, fl

ood

prot

ectio

n, a

nd p

hysi

cal s

ecur

ity

prot

ectio

n

Not

affe

cted

by

embe

dmen

t exc

ept f

or s

ecur

ity.

See

item

NH

SB 3

.1 fo

r phy

sica

l sec

urity

di

scus

sion

.

NH

SB-2

.1.6

The

NG

NP

shal

l be

desi

gned

to p

hysi

cally

pro

tect

the

safe

ty re

late

d SS

Cs

in th

e N

HSS

from

haz

ards

as

soci

ated

with

the

Hyd

roge

n Pr

oduc

tion

Syst

em, P

ower

C

onve

rsio

n Sy

stem

and

BO

P fa

cilit

ies.

Not

affe

cted

sig

nific

antly

by

embe

dmen

t. F

urth

er

anal

ysis

will

be re

quire

d in

con

cept

ual d

esig

n.Se

e se

ctio

n B1

.10.

NH

SB-2

.2R

esis

t all

stru

ctur

al lo

ads

as re

quire

d to

sup

port

safe

ty

func

tions

allo

cate

d to

the

NH

SB

Parti

al e

mbe

dmen

t is

favo

red

for r

educ

ed

seis

mic

resp

onse

for u

p to

50

perc

ent

embe

dmen

t. S

ee s

ectio

n B1

.9.

NH

SB-2

.2.1

.Pr

ovid

e st

ruct

ural

sup

port

for t

he re

acto

r pre

ssur

e ve

ssel

an

d its

inte

rnal

s Pa

rtial

em

bedm

ent i

s fa

vore

d fo

r red

uced

se

ism

ic re

spon

se fo

r up

to 5

0 pe

rcen

t em

bedm

ent.

See

sec

tion

B1.9

.

NH

SB-2

.2.2

.Pr

ovid

e st

ruct

ural

sup

port

for t

he P

HTS

Hel

ium

Pre

ssur

e Bo

unda

ry (H

PB) a

nd th

e pa

rt of

the

SHTS

HPB

insi

de

the

NH

SB

Parti

al e

mbe

dmen

t is

favo

red

for r

educ

ed

seis

mic

resp

onse

for u

p to

50

perc

ent

embe

dmen

t. S

ee s

ectio

n B1

.9.

NH

SB-2

.2.3

.Pr

ovid

e st

ruct

ural

sup

port

and

prot

ectio

n fo

r the

NSS

S SS

Cs

that

con

tain

sou

rces

of r

adio

activ

e m

ater

ial

outs

ide

the

reac

tor v

esse

l inc

ludi

ng th

e FH

SS, H

SS, a

nd

syst

ems

cont

aini

ng ra

dioa

ctiv

e w

aste

Parti

al e

mbe

dmen

t is

favo

red

for r

educ

ed

seis

mic

resp

onse

for u

p to

50

perc

ent

embe

dmen

t. S

ee s

ectio

n B1

.9.

NG

NP-

NH

S 10

0-R

XB

LDG

Rev

0N

GN

P C

once

ptua

l Des

ign

Stud

y Rea

ctor

Bui

ldin

g Fu

nctio

nal a

nd T

echn

ical

R

equi

rem

ents

and

Eva

luat

ion

of R

eact

or E

mbe

dmen

t

NG

NP-

NH

S10

0-RX

BLD

G0

Fina

l09

1608

.doc

Sept

embe

r 15,

200

817

0 of

186

T&FR

Num

ber

Req

uire

men

t R

ole

of R

eact

or E

mbe

dmen

t

NH

SB-2

.2.4

.Pr

otec

t the

NH

SB, P

HTS

HPB

and

oth

er N

HSS

SSC

s ag

ains

t loa

ds im

pose

d by

faul

ts in

clud

ing

pipe

bre

aks,

ch

emic

al re

leas

es, f

ires,

sei

smic

failu

res,

and

exp

losi

ons

loca

ted

in th

e H

PS A

cid

Dec

ompo

ser B

uild

ing,

SG

Bu

ildin

g, a

nd o

ther

adj

acen

t bui

ldin

gs a

ssoc

iate

d w

ith

the

HPS

, PC

S, a

nd b

alan

ce o

f pla

nt.

Prot

ect t

he P

HTS

H

PB fr

om lo

ads

impo

sed

by S

HTS

pip

ing

resu

lting

from

fa

ults

incl

udin

g st

ruct

ural

failu

res

in th

e H

PS A

cid

Dec

ompo

ser b

uild

ing

and

SG B

uild

ing

Not

affe

cted

sig

nific

antly

by

embe

dmen

t. S

ee

sect

ion

B1.1

0. A

naly

sis

will

be re

quire

d pe

r RG

1.

78 a

nd R

G 1

.91.

NH

SB-2

.2.5

.Pr

ovid

e st

ruct

ural

sup

port

for a

ll N

HSS

SSC

s th

atpr

ovid

e a

requ

ired

safe

ty fu

nctio

n fo

r all

sour

ces

of

radi

oact

ive

mat

eria

l with

in th

e N

HSB

Parti

al e

mbe

dmen

t is

favo

red

for r

educ

ed

seis

mic

resp

onse

for u

p to

50

perc

ent

embe

dmen

t. S

ee s

ectio

n B1

.9.

N

HSB

-2.2

.5.1

.Pr

ovid

e st

ruct

ural

sup

port

for S

SCs

requ

ired

for

conf

inem

ent o

f rad

ioac

tive

mat

eria

l Pa

rtial

em

bedm

ent i

s fa

vore

d fo

r red

uced

se

ism

ic re

spon

se fo

r up

to 5

0 pe

rcen

t em

bedm

ent.

See

sec

tion

B1.9

.

N

HSB

-2.2

.5.2

.Pr

ovid

e st

ruct

ural

sup

port

for S

SCs

requ

ired

to m

aint

ain

core

and

reac

tor p

ress

ure

vess

el g

eom

etry

Parti

al e

mbe

dmen

t is

favo

red

for r

educ

ed

seis

mic

resp

onse

for u

p to

50

perc

ent

embe

dmen

t. S

ee s

ectio

n B1

.9.

N

HSB

-2.2

.5.3

.Pr

ovid

e st

ruct

ural

sup

port

for S

SCs

requ

ired

to c

ontro

l co

re h

eat r

emov

al in

clud

ing

the

core

, rea

ctor

ves

sel,

reac

tor c

avity

and

RC

CS

Parti

al e

mbe

dmen

t is

favo

red

for r

educ

ed

seis

mic

resp

onse

for u

p to

50

perc

ent

embe

dmen

t. S

ee s

ectio

n B1

.9.

N

HSB

-2.2

.5.4

.Pr

ovid

e st

ruct

ural

sup

port

for S

SCs

requ

ired

for c

ontro

l of

hea

t gen

erat

ion

incl

udin

g th

e re

activ

ity c

ontro

l rod

sPa

rtial

em

bedm

ent i

s fa

vore

d fo

r red

uced

se

ism

ic re

spon

se fo

r up

to 5

0 pe

rcen

t em

bedm

ent.

See

sec

tion

B1.9

.

N

HSB

-2.2

.5.5

.Pr

ovid

e st

ruct

ural

sup

port

for S

SCs

requ

ired

for c

ontro

l of

che

mic

al a

ttack

Pa

rtial

em

bedm

ent i

s fa

vore

d fo

r red

uced

se

ism

ic re

spon

se fo

r up

to 5

0 pe

rcen

t em

bedm

ent.

See

sec

tion

B1.9

.

NG

NP-

NH

S 10

0-R

XB

LDG

Rev

0N

GN

P C

once

ptua

l Des

ign

Stud

y Rea

ctor

Bui

ldin

g Fu

nctio

nal a

nd T

echn

ical

R

equi

rem

ents

and

Eva

luat

ion

of R

eact

or E

mbe

dmen

t

NG

NP-

NH

S10

0-RX

BLD

G0

Fina

l09

1608

.doc

Sept

embe

r 15,

200

817

1 of

186

T&FR

Num

ber

Req

uire

men

t R

ole

of R

eact

or E

mbe

dmen

t

NH

SB-2

.2.6

.Pr

ovid

e st

ruct

ural

sup

port

for a

ll N

HSS

SSC

s th

at

prov

ide

a su

ppor

tive

safe

ty fu

nctio

n fo

r all

iden

tifie

d LB

Es a

s ne

cess

ary

to m

eet t

he T

LRC

(sup

porti

ve s

afet

y fu

nctio

n)

Parti

al e

mbe

dmen

t is

favo

red

for r

educ

ed

seis

mic

resp

onse

for u

p to

50

perc

ent

embe

dmen

t. S

ee s

ectio

n B1

.9.

NH

SB

-2.2

.7.

Prov

ide

conf

inem

ent o

f rad

ioac

tive

mat

eria

l dur

ing

norm

al o

pera

tion

and

durin

g LB

Es (A

OO

s, D

BEs,

and

BD

BEs)

as

nece

ssar

y to

mee

t the

TLR

C.

This

func

tion

is a

lso

supp

orte

d by

the

NH

SB P

RS

and

NH

SS H

VAC

(s

uppo

rtive

saf

ety

func

tion)

(A d

esig

n go

al o

f a s

ourc

e te

rm re

duct

ion

fact

or o

f 10

is s

et in

Sec

tion

A3 fo

r I-1

31an

d C

s-13

7 fo

r rel

ease

s fro

m th

e PH

TS d

urin

g de

sign

ba

sis

even

ts in

volv

ing

DLO

FC)

Leak

tigh

tnes

s in

not

affe

cted

by

leve

l of

embe

dmen

t. M

ost s

igni

fican

t lea

k pa

ths

are

door

s an

d H

VAC

dam

pers

and

val

ves.

The

se

will

requ

ire le

ak ti

ghtn

ess

rega

rdle

ss o

f em

bedm

ent l

evel

. Em

bedm

ent d

oes

not

cont

ribut

e si

gnifi

cant

ly to

leak

tigh

tnes

s.

NH

SB-2

.2.8

.C

ontro

l bui

ldin

g le

akag

e an

d lim

it ai

r ing

ress

to th

e co

re

follo

win

g a

larg

e br

each

or b

reac

hes

in th

e PH

TS H

PB,

FHSS

, and

HSS

for a

ll id

entif

ied

LBEs

as

nece

ssar

y to

m

eet t

he T

LRC

(sup

porti

ve s

afet

y fu

nctio

n)

(qua

ntifi

catio

n of

this

requ

irem

ent t

o be

det

erm

ined

)

Leak

tigh

tnes

s in

not

affe

cted

by

leve

l of

embe

dmen

t. M

ost s

igni

fican

t lea

k pa

ths

are

door

s an

d H

VAC

dam

pers

and

val

ves.

The

se

will

requ

ire le

ak ti

ghtn

ess

rega

rdle

ss o

f em

bedm

ent l

evel

. Em

bedm

ent d

oes

not

cont

ribut

e si

gnifi

cant

ly to

leak

tigh

tnes

s.

N

HSB

-2.3

Prot

ect t

he S

SCs

with

in th

e N

HSS

that

per

form

saf

ety

func

tions

from

all

inte

rnal

and

ext

erna

l haz

ards

as

iden

tifie

d in

the

LBEs

as

nece

ssar

y to

mee

t the

TLR

C

This

requ

irem

ent f

avor

s fu

ll em

bedm

ent.

How

ever

, pro

tect

ion

is fe

asib

le a

nd c

omm

on fo

r st

ruct

ures

with

out f

ull e

mbe

dmen

t. S

ee S

ectio

n B1

.8 N

atur

al P

heno

men

on H

azar

ds

NH

SB-2

.3.1

.Pr

otec

t the

SSC

s w

ithin

the

NH

SS th

at p

erfo

rm re

quire

d sa

fety

func

tions

from

LBE

s in

volv

ing

PHTS

and

SH

TS

HPB

leak

s an

d br

eaks

. Th

is fu

nctio

n is

als

o su

ppor

ted

by th

e N

HSS

Pre

ssur

e R

elie

f Sys

tem

.

Not

affe

cted

by

embe

dmen

t. S

ee s

ectio

n B1

.1

NH

SB-2

.3.2

.Pr

otec

t the

SSC

s w

ithin

the

NH

SS th

at p

erfo

rm re

quire

d sa

fety

func

tions

from

LBE

s in

volv

ing

mis

sile

s fro

m

rota

ting

mac

hine

ry a

nd o

ther

inte

rnal

sou

rces

Not

affe

cted

by

embe

dmen

t.

NG

NP-

NH

S 10

0-R

XB

LDG

Rev

0N

GN

P C

once

ptua

l Des

ign

Stud

y Rea

ctor

Bui

ldin

g Fu

nctio

nal a

nd T

echn

ical

R

equi

rem

ents

and

Eva

luat

ion

of R

eact

or E

mbe

dmen

t

NG

NP-

NH

S10

0-RX

BLD

G0

Fina

l09

1608

.doc

Sept

embe

r 15,

200

817

2 of

186

T&FR

Num

ber

Req

uire

men

t R

ole

of R

eact

or E

mbe

dmen

t

NH

SB-2

.3.3

.Pr

otec

t the

SSC

s w

ithin

the

NH

SS th

at p

erfo

rm re

quire

d sa

fety

func

tions

from

LB

Es

invo

lvin

g in

tern

al fi

res

Not

affe

cted

by

embe

dmen

t.

NH

SB-2

.3.4

.Pr

otec

t the

SSC

s w

ithin

the

NH

SS th

at p

erfo

rm re

quire

d sa

fety

func

tions

from

LBE

s in

volv

ing

inte

rnal

floo

dsN

ot a

ffect

ed b

y em

bedm

ent.

NH

SB-2

.3.5

.Pr

otec

t the

SSC

s w

ithin

the

NH

SS th

at p

erfo

rm re

quire

d sa

fety

func

tions

from

LB

Es

invo

lvin

g hy

drog

en p

roce

ss

haza

rds

Not

affe

cted

sig

nific

antly

by

embe

dmen

t. S

ee

sect

ion

B1.1

0. A

naly

sis

will

be re

quire

d pe

r RG

1.

78 a

nd R

G 1

.91.

NH

SB-2

.3.6

.Pr

otec

t the

SSC

s w

ithin

the

NH

SS th

at p

erfo

rm re

quire

d sa

fety

func

tions

from

LBE

s in

volv

ing

seis

mic

eve

nts

Parti

al e

mbe

dmen

t is

favo

red

for r

educ

ed

seis

mic

resp

onse

for u

p to

50

perc

ent

embe

dmen

t. S

ee s

ectio

n B1

.9.

NH

SB-2

.3.7

.Pr

otec

t the

SSC

s w

ithin

the

NH

SS th

at p

erfo

rm re

quire

d sa

fety

func

tions

from

LBE

s in

volv

ing

acci

dent

airc

raft

cras

hes

and

trans

porta

tion

acci

dent

s

This

requ

irem

ent f

avor

s fu

ll em

bedm

ent.

How

ever

, pro

tect

ion

is fe

asib

le a

nd c

omm

on fo

r st

ruct

ures

with

out f

ull e

mbe

dmen

t. S

ee S

ectio

n B1

.7 M

alev

olen

t Haz

ards

NH

SB-2

.3.8

.Pr

otec

t the

SSC

s w

ithin

the

NH

SS th

at p

erfo

rm re

quire

dsa

fety

func

tions

from

LBE

s in

volv

ing

high

win

ds a

nd

win

d ge

nera

ted

mis

sile

s

This

requ

irem

ent f

avor

s fu

ll em

bedm

ent.

How

ever

, pro

tect

ion

is fe

asib

le a

nd c

omm

on fo

r st

ruct

ures

with

out f

ull e

mbe

dmen

t. S

ee S

ectio

n B1

.8 N

atur

al P

heno

men

on H

azar

ds

N

HSB

-2.3

.9.

Prot

ect t

he S

SCs

with

in th

e N

HSS

that

per

form

requ

ired

safe

ty fu

nctio

ns fr

om L

BEs

invo

lvin

g ot

her i

nter

nal a

nd

exte

rnal

haz

ards

(req

uire

men

ts to

be

dete

rmin

ed)

This

requ

irem

ent f

avor

s fu

ll em

bedm

ent.

How

ever

, pro

tect

ion

is fe

asib

le a

nd c

omm

on fo

r st

ruct

ures

with

out f

ull e

mbe

dmen

t. S

ee S

ectio

n B1

.8 N

atur

al P

heno

men

on H

azar

ds

NH

SB-2

.3.1

0.Pr

otec

t the

SSC

s w

ithin

the

NH

SS th

at p

erfo

rm

supp

ortiv

e sa

fety

func

tions

from

LBE

s as

nec

essa

ry to

m

eet t

he to

p le

vel r

egul

ator

y cr

iteria

as

nece

ssar

y to

m

eet t

he T

LRC

(req

uire

men

ts to

be

dete

rmin

ed)

Not

affe

cted

by

embe

dmen

t. S

ee s

ectio

n B1

.1

N

HSB

-2.4

Prov

ide

phys

ical

sec

urity

of v

ital a

reas

with

in th

e N

HSB

ag

ains

t act

s of

sab

otag

e an

d te

rroris

mTh

is re

quire

men

t fav

ors

full

embe

dmen

t.H

owev

er, p

rote

ctio

n is

feas

ible

and

com

mon

for

NG

NP-

NH

S 10

0-R

XB

LDG

Rev

0N

GN

P C

once

ptua

l Des

ign

Stud

y Rea

ctor

Bui

ldin

g Fu

nctio

nal a

nd T

echn

ical

R

equi

rem

ents

and

Eva

luat

ion

of R

eact

or E

mbe

dmen

t

NG

NP-

NH

S10

0-RX

BLD

G0

Fina

l09

1608

.doc

Sept

embe

r 15,

200

817

3 of

186

T&FR

Num

ber

Req

uire

men

t R

ole

of R

eact

or E

mbe

dmen

t

stru

ctur

es w

ithou

t ful

l em

bedm

ent.

See

Sec

tion

B1.7

Mal

evol

ent H

azar

ds

NH

SB-3

.0Th

e N

HSB

Pre

ssur

e R

elie

f Sys

tem

sha

ll pe

rform

the

follo

win

g fu

nctio

ns:

N

HSB

-3.1

The

PRS

shal

l be

com

patib

le w

ith th

e N

HSB

bou

ndar

yfu

nctio

nal r

equi

rem

ent.

Not

App

licab

le th

is is

a G

ener

al R

equi

rem

ent

N

HSB

-3.2

The

PRS

shal

l be

desi

gned

, and

the

NH

SB

com

partm

ents

sha

ll be

siz

ed s

o th

at le

aks

and

brea

ks u

p to

10

mm

equ

ival

ent b

reak

siz

e on

the

PHTS

HPB

or

that

par

t of t

he S

HTS

HPB

insi

de th

e N

HSB

do

not o

pen

the

PRS

so th

at H

VAC

filtr

atio

n ca

pabi

lity

shal

l be

cont

inuo

usly

mai

ntai

ned

Not

affe

cted

by

embe

dmen

t. It

is e

xpec

ted

that

PR

S w

ill be

ven

ted

to a

rele

ase

poin

t at t

he to

p th

e bu

ildin

g w

ith o

r with

out f

ilter

s.

N

HSB

-3.3

PRS

shal

l ope

n to

pre

vent

ove

rpre

ssur

e an

d th

erm

al

dam

age

to a

ny s

afet

y re

late

d SS

C, r

emov

e th

e pr

essu

re

driv

ing

forc

e fo

r rad

ionu

clid

e re

leas

es fr

om th

e re

acto

r bu

ildin

g, a

nd d

epen

ding

on

the

actu

al d

esig

n th

at is

ch

osen

, may

als

o be

requ

ired

to re

clos

e to

ena

ble

post

-bl

ow-d

own

filtra

tion

for t

he fo

llow

ing

desi

gn b

asis

eve

nt

cond

ition

s.

Not

affe

cted

by

embe

dmen

t. It

is e

xpec

ted

that

PR

S w

ill be

ven

ted

to a

rele

ase

poin

t at t

he to

p th

e bu

ildin

g w

ith o

r with

out f

ilter

s.

NH

SB-3

.3.1

.Br

eaks

in th

e PH

TS H

PB in

eac

h N

HSB

com

partm

ent

cont

aini

ng N

HSS

pip

ing

or v

esse

ls u

p to

100

mm

equi

vale

nt s

ingl

e en

ded

brea

k si

ze

Not

affe

cted

by

embe

dmen

t. It

is e

xpec

ted

that

PRS

will

be v

ente

d to

a re

leas

e po

int a

t the

top

the

build

ing

with

or w

ithou

t filt

ers.

NH

SB-3

.3.2

.Br

eaks

in th

e SH

TS H

PB in

eac

h N

HSB

com

partm

ent

cont

aini

ng N

HSS

pip

ing

or v

esse

ls u

p to

1m

eter

equi

vale

nt s

ingl

e en

ded

brea

k si

ze

Not

affe

cted

by

embe

dmen

t

NH

SB-3

.3.3

Brea

ks in

the

HSS

pip

ing

up to

100

mm

equi

vale

ntsi

ngle

end

ed b

reak

siz

eN

ot a

ffect

ed b

y em

bedm

ent

NG

NP-

NH

S 10

0-R

XB

LDG

Rev

0N

GN

P C

once

ptua

l Des

ign

Stud

y Rea

ctor

Bui

ldin

g Fu

nctio

nal a

nd T

echn

ical

R

equi

rem

ents

and

Eva

luat

ion

of R

eact

or E

mbe

dmen

t

NG

NP-

NH

S10

0-RX

BLD

G0

Fina

l09

1608

.doc

Sept

embe

r 15,

200

817

4 of

186

T&FR

Num

ber

Req

uire

men

t R

ole

of R

eact

or E

mbe

dmen

t

N

HSB

-3.4

The

PRS

shal

l ope

n an

d pr

even

t exc

essi

ve d

amag

e to

th

e N

HSB

in re

spon

se to

a B

eyon

d D

esig

n Ba

sis

1m

eter

bre

ak in

the

PHTS

pip

ing;

exc

essi

ve d

amag

e is

de

fined

by

exce

edin

g th

e TL

RC

usi

ng re

alis

tic

assu

mpt

ions

Not

affe

cted

by

embe

dmen

t

NH

SB-4

.0Th

e N

HSS

HVA

C S

yste

m s

hall

perfo

rm th

e fo

llow

ing

func

tions

:

N

HSB

-4.1

a)M

aint

ain

inte

rnal

env

ironm

enta

l con

ditio

ns fo

rall

NH

SS S

SCs

durin

g no

rmal

ope

ratio

n,b)

sur

vive

con

ditio

ns fo

r all

AOO

and

DBE

LBE

s, a

ndc)

pro

vide

a p

ost-e

vent

cle

anup

func

tion.

(sup

porti

ve

safe

ty fu

nctio

n)

It co

uld

be s

pecu

late

d th

at e

mbe

dmen

t mig

ht

redu

ce H

VAC

sys

tem

load

s w

ith re

duct

ion

of

sola

r and

tran

smis

sion

hea

t loa

ds.

HVA

C

syst

em s

izin

g an

d pe

rform

ance

requ

irem

ents

are

no

t usu

ally

driv

en s

igni

fican

tly b

y ex

tern

al

cond

ition

s. C

oncr

ete

stru

ctur

es h

ave

subs

tant

ial

ther

mal

iner

tial t

o da

mpe

n ef

fect

s of

ext

erna

l te

mpe

ratu

res.

HVA

C fl

ows

are

driv

en m

ore

by

inte

rnal

load

s an

d co

ntam

inat

ion

cont

rol n

eeds

.

N

HSB

-4.2

Mai

ntai

n en

viro

nmen

tal c

ondi

tions

for S

SCs

that

per

form

su

ppor

tive

safe

ty fu

nctio

ns d

urin

g LB

Es (s

uppo

rtive

sa

fety

func

tion)

Not

affe

cted

by

embe

dmen

t

N

HSB

-4.3

Perfo

rm ra

dion

uclid

e fil

tratio

n fu

nctio

ns a

s re

quire

d to

m

eet t

he T

LRC

for a

ll LB

Es (s

uppo

rtive

saf

ety

func

tion)

Not

affe

cted

by

embe

dmen

t

N

HSB

-4.4

Perfo

rm p

rote

ctiv

e fu

nctio

ns to

isol

ate

the

HVA

C d

urin

g D

BE a

nd B

DBE

HPB

bre

aks

to p

reve

nt d

amag

e fro

m

high

tem

pera

ture

s an

d pr

essu

res

in b

uild

ing

com

partm

ents

dur

ing

depr

essu

rizat

ion

to e

nabl

e po

st

blow

-dow

n fil

tratio

n (s

uppo

rtive

saf

ety

func

tion)

Not

affe

cted

by

embe

dmen

t




B2.7 RECOMMENDATIONS

The partial embedment scheme scores best for both Rock site and Soil site, and istherefore be recommended. The degree of partial embedment will be optimized for the site during the conceptual design phase. The case of soil site and a deep water table is not shown in ranking table but it is expected the scores would be essentially the same as for a rock site with deep water table.

The PBMR design is quite flexible to accommodate varying site geotechnical conditions.Access to the reactor building can be adjusted for a given site without a significant impact on the layout of major systems within the building.




B3 OPEN ISSUES AND ADDITIONAL R&D AND ENGINEERING STUDIES

1. Analysis will be required for quantities of explosive or flammable gasses and materials at or near the site including Hydrogen in order to meet requirements of Regulatory Guide 1.91. The intent will be to limit these quantities and maintain safe distances as required to prevent any substantial damage to the safety related nuclear heat supply system SSCs.

2. Review of life cycle costs associated with operations, inspections, maintenance, and energy costs, associated with HVAC loads, is recommended for the conceptual design.

3. Preparation of a Seismic Basis of Design Document is expected early in Conceptual Design to implement methodology promulgated by the current regulations.

4. There no R&D items identified in this portion of the study.




REFERENCES

[1] NGNP and Hydrogen Production Conceptual Design Report, Section 2 User Requirements, NGNP-02-RPT-001, April 2007

[2] NGNP and Hydrogen Production Conceptual Design Report, Section 14 Safety, NGNP-14-RPT-001, April 2007

[3] PBMR (Pty) Ltd., ‘U.S. Design Certification – Probabilistic Risk Assessment Approach for the Pebble Bed Modular Reactor,’ June 13, 2006.

[4] PBMR (Pty) Ltd., ‘U.S. Design Certification – Licensing Basis Event Selection for the Pebble Bed Modular Reactor,’ June 30, 2006.

[5] PBMR (Pty) Ltd., ‘U.S. Design Certification – Safety Classification of Structures, Systems, and Components for the Pebble Bed Modular Reactor,’ June 30, 2006.

[6] PBMR (Pty) Ltd., ‘U.S. Design Certification – Defense-in-Depth Approach for the Pebble Bed Modular Reactor,’ August, 2006.

[7] U.S. Nuclear Regulatory Commission, NUREG-1860, “Feasibility Study for a Risk-Informed and Performance Based Regulatory Structure for Future Plant Licensing”, December 2007

[8] U.S. Code of Federal Regulations, Parts 50.34 and 50.67[9] Manual of Protective Action Guides and Protective Actions for Nuclear Incidents,

EPA/400/R-92/001, May 1992.[10] NGNP and Hydrogen Production Preconceptual Design Study - Contamination Control

(NGNP-NHS 50-CC, Rev. A, April 2008[11] PBMR Calculated Fuel Release Source Terms for 500Mwt Design with Different

Operating Temperatures than NGNP (Proprietary Data)[12] GT MHR Preliminary Safety Assessment Report (DOE-GT-MHR-100230, Rev. 0,

September 1994)”, Table 4.4.3-3[13] Preliminary Safety Information document for the Standard MHTGR and responses to

NRC Comments”, DOE-HTGR-86-024, Amendment 13, August 7, 1992[14] PBMR Recommended Values for Dust Re-suspension during DLOFC (Proprietary

Data)[15] IAEA TECDOC-978, “Fuel Performance and Fission Product Behavior in Gas Cooled

Reactors”, Nov. 1997[16] NRC Reg. Guide 1.183, July 2000[17] Federal Guidance Reports 11 (1988) and 12 (1993)[18] NGNP-04-DWG-001-REVB-Nuclear Heat Supply Building, April 2007[19] NGNP-04-DWG-002-REVB-Nuclear Heat Supply Building Floor Level -9250, April

2007[20] NGNP-04-DWG-003-REVB-Nuclear Heat Supply Building Floor Level + 700, April

2007[21] NGNP-04-DWG-004-REVB-Nuclear Heat Supply Building Floor Level +16550, April

2007[22] Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion

Factors for Inhalation, Submersion, and Ingestion, Federal Guidance Report 11, EPA-520/1-88-020, Environmental Protection Agency, 1988.




[23] External Exposure to Radio-nuclides in Air, Water, and Soil, Federal Guidance Report 12, EPA-402-R-93-081, Environmental Protection Agency, 1993.

[24] Electric Power Research Institute (EPRI)’s Advanced Light Water Reactor UtilityRequirements Document , Revision 8

[25] NEI 03-12, [Site] Physical Security Plan, Training & Qualification Plan and Safeguards Contingency Plan.

[26] Methodology of Performing Aircraft Impact Assessment for New Plant Designs soon to be issued as NEI 08-13.

[27] DLOFC Transient Analysis of RCCS Standpipes, PBMR Calculation TA00024/A issued 5 June 2004. (Proprietary Data)

[28] Heat Transfer Calculation for Reactor Embedment Task Design Information Transmittal DIT001254 taking information from Calculation T001284 August 2008.(Proprietary Data)

[29] Electric Power Research Institute (EPRI), “Guidelines for Permanent BWR Hydrogen Water Chemistry Installations – 1987 Revision”. Report No. EPRI NP-5283-SR-a,September 1987.

[30] NIOSH, NIOSH Pocket Guide to Chemical Hazards, National Institute for Occupational Safety and Health, 1997.

[31] Ramsdell, J. V. Jr. and C. A. Simonen, "Atmospheric Relative Concentrations in Building Wakes". Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, PNL-10521, NUREG/CR-6331, Rev. 1, May 1997.

[32] Smith, Curtis, Scott Beck, and Bill Galyean, “An Engineering Analysis for Separation Requirements of a Hydrogen Production Plant and High-Temperature Nuclear Reactor”. Prepared by Idaho National Laboratory, Report No. INL/EXT-05-0137, Rev.1, July 2006.

[33] U. S. Environmental Protection Agency, ALOHA User’s Manual, National Oceanic and Atmospheric Administration, Version 5.4.1, February 2007.

[34] U. S. Nuclear Regulatory Commission, NUREG/CR-6944, “Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs), Volume 6: “Process Heat and Hydrogen Co-generation PIRTs”, March 2008.

[35] U. S. Nuclear Regulatory Commission, Regulatory Guide 1.78, Evaluating the Habitability of a Nuclear Power Plant Control Room during a Postulated HazardousChemical Release, Revision 1, December 2001.

[36] U. S. Nuclear Regulatory Commission, Regulatory Guide 1.91, “Evaluations of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants”, Revision1, February 1978.

[37] Postma, M. and Carl Walker, Inc. (2006), Building Envelope Design Guide -Foundation Walls, Whole Building Design Guide, Building Enclosure Council

[38] CETCO Liquid Boot Company, www.liquidboot.com.wp.specifications.php[39] Stone & Webster Engineering Corporation (1979), Technical Guideline GTG-6.15-1,

“Determination of Lateral Pressures on Buried Structures in Granular Soils”[40] “Ground Motion Considerations for Nuclear Power Plant Design with Emphasis on

Soil-Structural Interaction Aspects” by R. P. Kennedy, M.S. Power, and C. Y. Chang,




presented in the Proceedings of the Workshop on Soil-Structure Interaction, Bethesda, Maryland, June 16-18, 1986.

[41] ASCE Standard 4-98, "Seismic Analysis of Safety-Related Nuclear Structures and Commentary," American Society of Civil Engineers, September 1998.

[42] ASTM D 5321-02, Standard Test Method for Determining the Coefficient of Soil and Geosynthetic or Geosynthetic and Geosynthetic Friction by the Direct Shear Method, ASTM International, West Conshohoken, PA.

[43] Coduto, D. P. (2001), Foundation Design Principles and Practices, 2nd Edition, Prentice Hall.

[44] EduPro (1999). EduPro Civil Systems, Inc., ProShake User’s Manual Version 1.1, June.

[45] U.S Nuclear Regulatory Commission (2007), Standard Review Plan 3.7.2, Seismic System Analysis, Rev. 3, March.

[46] U.S. Nuclear Regulatory Commission (2007), Standard Review Plan 2.5.1, Basic Geologic and Seismic Information, Rev. 4, March.

[47] U.S. Department of Energy (2002), DOE-STD-1022-94, Natural Phenomena Hazardsand Site Characterization Criteria

[48] Karl N. Fleming, John Fletcher, and Neil Broom, “RI-ISI Program For Modular HTGRs”, Proceedings HTR2006: 3rd International Topical Meeting on High Temperature Reactor Technology, October 1-4, 2006, Johannesburg, South Africa

[49] Karl Fleming, Steve Gosselin, and Ron Gamble, “Reliability and Integrity Management (RIM) Program For Modular High Temperature Gas-Cooled Reactors (MHRs) -Technical Basis Document for ASME Section XI Division 2 for MHRs, ASME Section XI Section Special Working Group on HTGRs, June 2008

[50] NRC Regulatory Guide 1.4, Assumptions used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors, Revision 2, June 1974.

[51] Occupational Safety and Health Standards, 29 CFR 1910.103, “Hydrogen” (Hazardous Materials), Rev. 72 FR 71069, Dec. 14, 2007

[52] Occupational Safety and Health Standards,29 CFR 1910.104, “Oxygen” (Hazardous Materials), Rev. 61 FR 9237, March 7, 1996

[53] Occupational Safety and Health Standards, 29 CFR 1910.119, “Process safety management of highly hazardous chemicals” (Hazardous Materials), Rev. 61 FR 9227, March 7, 1996

[54] Nuclear Regulatory Commission, 10 CFR 50, “Domestic Licensing of Production and Utilization Facilities”, Rev. 73 FR 42674, Jul. 23, 2008

[55] Nuclear Regulatory Commission, 10 CFR 52, “Early Site Permits; Standard Design Certifications and Combined Licenses for Nuclear Power Plants”, Rev. 54 FR 15386, Apr. 18, 1989

[56] Nuclear Regulatory Commission,10 CFR 73, “Physical Protection of Plants andMaterials”, Rev. 38 FR 35430, Dec. 28, 1973

[57] Nuclear Regulatory Commission, “Reactor Site Criteria”,10 CFR 100, Rev. 27 FR 3509, April 1962




[58] American Society for Testing and Materials, ASTM 1557 “Standard Test Methods for Laboratory Compaction Characteristics of Soil using Modified Effort”, Nov. 1,2007

[59] Occupational Safety and Health Standards, 29 CFR 1926 “Safety and Health Regulations for Construction”, Rev. 71 FR 16675, April 3, 2006

[60] Nuclear Regulatory Commission, 10 CFR 50.54,“Requirements for Renewal ofOperating Licenses for Nuclear Power Plants”, Rev. 60 FR 22491, May 8, 1995

[61] ASCE Standard 7-05 Standard, “Minimum Design Loads for Buildings and Other Structures”, Rev. 0, Jan. 1,2006

[62] Nuclear Regulatory Commission, Regulatory Guide 1.76“Design-basis Tornado and Tornado Missiles for Nuclear Power Plants”, March 2007

[63] Nuclear Regulatory Commission, Regulatory Guide 1.60 “Design Response Spectra for Seismic Design of Nuclear Power Plants” December 1973

[64] Concrete Nuclear Structures, ACI 349-06 “Code Requirements for Nuclear Safety-Related Concrete Structures”, Rev. 0, Sept. 1, 2007

[65] AISC Committee, AISC N690L-03 “Load and Resistance Factor Design Specification for Safety-Related Steel Structures for Nuclear Facilities”, Rev. 1, Dec. 17, 2003

[66] International Code Council, IBC-06, “International Building Code”, 2006[67] ACI 318-06, Concrete Nuclear Structures, “Building Code Requirements for Structural

Concrete and Commentary”, 2005[68] Nuclear Regulatory Commission, NUREG 6210 “HABIT, Toxic and Radioactive

Release Hazards in Reactor Control Room”, Nov. 7, 2005[69] Calculation 132155.CAL-C-0001, Rev.0, NGNP Reactor Building Embedment Cost

Analysis, September 2008[70] Nuclear Regulatory Commission, Regulatory Guide, 1.29 “Seismic Design

Classification” September 1978[71] Nuclear Regulatory Commission, Regulatory Guide 1.165, “Identification and

Characterization of Seismic Sources and Determination of Safe Shutdown Earthquake Ground Motion” March 1997

[72] Nuclear Regulatory Commission, 10 CFR 100.23, "Geologic and Seismic Siting Criteria,"

[73] Nuclear Regulatory Commission, Regulatory Guide 1.208, “A Performance-basedApproach to Define the Site-Specific Earthquake Ground Motion” March 2007

[74] ASCE Standard ASCE/SEI 43-05, "Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities," 2005

[75] Department of Energy Standard 1020, “Natural Phenomena Hazards Design and Evaluation Criteria for Department of Energy Facilities”, January 2002

[76] Nuclear Regulatory Commission, 10 CFR Part 50, Appendix S, “Earthquake Engineering Criteria for Nuclear Power Plants”.

[77] AISC ASD/LRFD Steel Construction manual, 13th Edition[78] Nuclear Regulatory Commission, Draft Regulatory Guide DG-1146, “A Performance-

based Approach to Define the Site-Specific Earthquake Ground Motion” October 2006[79] Nuclear Regulatory Commission, 10 CFR Part 20, “Standards for Protection Against

Radiation”




BIBLIOGRAPHY

All references are provided in the references section of the report




DEFINITIONSNone




REQUIREMENTS

Not used




ASSUMPTIONS

1. For purposes of estimating cost, the reactor building is assumed to be 65.8 m high measured from the top of concrete mat to the roof. Two assumed configurations of structures with approximately equal volumes were considered.

• A circular structure with a 56 m outside diameter

• A square structure with 50 m long exterior sides

2. Two foundation conditions were assumed to bound the potential site geological conditions. The first condition assumed a rock site with a water table below the bottom of the reactor building mat under full embedment of the structure. This condition is comparable to the condition expected at the INL site. The second condition is a deep competent soil site with a high water table. This condition would be anticipated along the southeastern coast and Gulf coast of the United States.

3. Table 20 identifies the major systems to be housed in the Reactor Building, the relative safety classification and requirements for location with respect to the Reactor and other systems. The following safety classifications were taken from Reference [2] for input to this table. These safety classifications are assumed for the purposes of this study only in order to gain an understanding of the portion of the Reactor Building that will house SSCs requiring protection from internal andexternal hazards. Table 20 is not to be interpreted as a formal position on the SSC safety classification. Safety classification will be developed during conceptual and preliminary design based on the RI-PB safety analysis and licensing approach.

Safety-Related SSCs (SR):

This category is for SSCs relied on to perform required safety functions to mitigate the public consequences of Design Basis Events (DBEs) to comply with the dose limits of 10 CFR §50.34[8]

This category is also for SSCs relied on to perform required safety functions to prevent the frequency of Beyond Design Basis Accidents (BDBEs) with consequences greater than the 10 CFR §50.34[8] dose limits from increasing into the DBE region.

Non-Safety-Related with Special Treatment (NSR-ST):

This category is for SSCs relied on to perform safety functions to mitigate the consequences of Anticipated Operational Occurrences (AOOs) to comply with the offsite dose limits of 10 CFR Part 20[79].




This category is also for SSCs relied on to perform safety functions to prevent the frequency of DBEs with consequences greater than the 10 CFR Part 20 [79]offsite dose limits from increasing into the AOO region.

Non-Safety-Related with No Special Treatment (NSR)

This category is for all SSCs not included in either of the above two categories.

4. For additional assumptions see “Key Assumptions” in section A3.3.




APPENDICES

APPENDIX A: 90 PERCENT DESIGN REVIEW PRESENTATION TO BEA

.

Slid

e1

PB

MR

TE

AM

Rea

ctor

Bui

ldin

g Fu

nctio

nal a

nd T

echn

ical

R

equi

rem

ents

90%

Des

ign

Rev

iew

Sept

embe

r 5, 2

008

Slid

e1

NG

NP

TE

AMEv

alua

tion

of R

eact

or E

mbe

dmen

t

90%

Des

ign

Rev

iew

Sept

embe

r 5, 2

008

Slid

e2

NG

NP

TE

AM

Obj

ectiv

es o

f Stu

dy

–R

evie

w is

sues

that

are

cru

cial

for d

eter

min

ing

the

reac

tor

embe

dmen

t dep

th

–Id

entif

y al

tern

ativ

e R

eact

or e

mbe

dmen

t con

figur

atio

ns a

pplic

able

to a

var

iety

of s

ite c

ondi

tions

incl

udin

g th

e IN

L si

te–

Dev

elop

an

obje

ctiv

e sc

orin

g m

etho

d to

eva

luat

e an

d ra

nk

alte

rnat

ives

with

resp

ect t

o si

te c

ondi

tions

and

em

bedm

ent f

or IN

L an

d fo

r soi

l site

s–

Mak

e re

com

men

datio

ns o

f the

pre

ferr

ed e

mbe

dmen

t con

figur

atio

n–

The

stud

y re

port

of R

eact

or E

mbe

dmen

t is

to b

e co

mbi

ned

as P

art

B w

ith th

e re

port

on

Rea

ctor

Bui

ldin

g Fu

nctio

ns a

nd T

echn

ical

R

equi

rem

ents

Par

t A. T

he re

latio

nshi

p of

the

two

part

s w

ill b

e di

scus

sed.

Slid

e3

NG

NP

TE

AM

Flow

Cha

rt

Slid

e4

NG

NP

TE

AM

Con

trib

utor

s

•P

BM

R: D

irk U

ys, R

oger

You

ng•

Wes

tingh

ouse

: Jim

Win

ters

•S

haw

: P

eter

Wel

ls, B

ob W

ilmer

, Will

iam

Mar

tin, M

ike

Dav

idso

n, R

icha

rd D

econ

to, S

teve

Vig

eant

, Pau

l S

latk

avitz

•Te

chno

logy

Insi

ghts

: Kar

l Fle

min

g ,

Slid

e5

NG

NP

TE

AM

Issu

es re

lativ

e to

Rea

ctor

Em

bedm

ent

•O

pera

tiona

l Nee

ds–

Des

ign

for O

pera

tiona

l effe

cts

(Rea

ctor

pro

tect

ion,

acc

ess

for r

efue

ling,

etc

.) fo

r no

rmal

and

em

erge

ncy

oper

atio

ns, D

BE

s, a

nd m

aint

enan

ce a

ctiv

ities

. Rel

ativ

e lo

catio

ns o

f sub

-sys

tem

s w

ith re

spec

t to

reac

tor a

re c

onsi

dere

d.•

Hea

t Dis

sipa

tion

to E

nviro

nmen

t–

The

abili

ty o

f the

bui

ldin

g to

tran

sfer

hea

t to

the

envi

ronm

ent e

ither

to g

roun

d or

to

am

bien

t air

will

be

asse

ssed

.•

Wat

er T

able

Effe

cts

–G

roun

dwat

er le

vels

are

eva

luat

ed w

ith re

gard

to im

pact

on

stru

ctur

al d

esig

n,

cons

truct

ion

dew

ater

ing

sche

mes

and

nee

d fo

r wat

erpr

oofin

g.•

Geo

tech

nica

l Con

stra

ints

and

Fou

ndat

ion

Perf

orm

ance

–A

rang

e of

foun

datio

n m

ater

ials

are

con

side

red

with

rega

rd to

sta

tic a

nd

dyna

mic

pro

perti

es a

nd fo

unda

tion

perfo

rman

ce.

•C

onst

ruct

ion

Con

side

ratio

ns–

The

effe

cts

of th

e em

bedm

ent d

epth

on

cons

truct

abili

ty u

sing

mod

ern

tech

niqu

es s

uch

as m

odul

ariz

atio

n ar

e co

nsid

ered

. The

exc

avat

ion

met

hod

used

will

requ

ire a

n ev

alua

tion

of s

ite s

pace

con

stra

ints

, and

cos

t and

sch

edul

e im

pact

s

Slid

e6

NG

NP

TE

AM

Issu

es re

lativ

e to

Rea

ctor

Em

bedm

ent (

cont

)•

Cos

t Ben

efit

–C

ost b

enef

it in

clud

ing

key

elem

ents

of r

elat

ive

capi

tal c

ost w

ill be

dev

elop

ed fo

r var

ious

em

bedm

ent o

ptio

ns a

nd c

ompa

red

in a

cos

ttab

le.

•M

alev

olen

t Haz

ards

–Th

is e

lem

ent a

ddre

sses

pro

tect

ion

agai

nst t

he d

esig

n ba

sis

safe

guar

ds a

nd s

ecur

ity th

reat

D

BT a

nd th

e be

yond

des

ign

thre

at im

pose

d by

a p

oten

tial n

on a

ccid

enta

l airc

raft

strik

e, o

r ot

her B

DBE

.•

Nat

ural

Haz

ards

–N

atur

al P

heno

men

on H

azar

ds (N

PH),

such

as

torn

ado

(incl

udin

g m

issi

les)

and

floo

ding

will

are

cons

ider

ed.

•G

eolo

gica

l Haz

ards

–Th

e po

tent

ial i

mpa

cts

that

nat

ural

geo

logi

cal h

azar

ds s

uch

as d

isso

lutio

n fe

atur

es in

sol

uble

ro

ck, w

eak

com

pres

sibl

e so

ils, s

lope

inst

abilit

y re

sulti

ng in

land

slid

e po

tent

ial,

and

lique

fact

ion

and

othe

r ear

thqu

ake

indu

ced

flow

phe

nom

ena

are

cons

ider

ed•

Che

mic

al R

elea

ses

–C

onsi

dera

tion

for t

he h

azar

ds o

f che

mic

als

stor

ed a

nd o

r tra

nspo

rted

in ra

nge

of th

e R

eact

or,

the

pote

ntia

l effe

cts,

and

miti

gatin

g fe

atur

es th

at c

ould

be

utiliz

ed. T

his

type

of c

once

rn d

oes

ofte

n m

ater

ializ

e w

ith o

pera

ting

faci

litie

s lo

cate

d of

act

ive

indu

stria

l are

as. T

he p

oten

tial f

ro

hydr

ogen

exp

losi

ons

is a

ddre

ssed

her

e.

Slid

e7

NG

NP

TE

AM

Alte

rnat

ive

Rea

ctor

Em

bedm

ent C

once

pts

•M

inim

al e

mbe

dmen

t of t

he b

uild

ing

7 -1

0 m

•Pa

rtia

l em

bedm

ent 2

0 -3

0 m

•

Full

Embe

dmen

t >60

m

•A

ltern

ativ

es to

be

asse

ssed

for r

ock

site

s w

ith d

eep

wat

er ta

ble

sim

ilar t

o th

e IN

L si

te a

nd s

oil s

ite w

ith

shal

low

wat

er ta

ble

Slid

e8

NG

NP

TE

AM

Ope

ratio

ns N

eeds

•P

rovi

de m

aint

enan

ce a

cces

s to

maj

or e

quip

men

t•

Pro

vide

saf

e pe

rson

nel a

cces

s an

d eg

ress

pat

hs (e

leva

tors

, cor

ridor

s an

d st

airw

ells

)•

Ena

ble

plac

emen

t of s

yste

m c

ompo

nent

s as

nee

ded

rela

tive

to th

e el

evat

ion

of th

e re

acto

r its

elf,

to m

eet t

herm

al h

ydra

ulic

and

oth

er o

pera

tiona

l nee

ds.

•P

rovi

de p

rote

ctio

n of

saf

ety

rela

ted

com

pone

nts

from

ext

erna

l haz

ards

and

haz

ards

due

to

inte

rnal

SS

C fa

ilure

s•

The

PB

MR

des

ign

does

not

requ

ire ro

utin

e ac

cess

to th

e R

eact

or h

ead

for r

efue

ling

activ

ities

. Thi

s is

sue

driv

es o

ther

HTG

R c

onfig

urat

ions

to fa

vorf

ull e

mbe

dmen

t.•

Em

bedm

ent c

ould

effe

ct th

e ar

rang

emen

t of t

he P

ress

ure

Rel

ief S

yste

m w

ith v

ent s

tack

if

requ

ired.

It is

exp

ecte

d th

at th

e P

RS

will

dis

char

ge fr

om th

eto

p of

the

build

ing.

All

off

site

dos

e es

timat

es u

sed

in th

is s

tudy

Par

t ass

ume

grou

nd le

vel r

elea

se.

•Th

e ge

nera

l fin

ding

is th

at e

mbe

dmen

t dec

isio

n is

not

driv

en s

igni

fican

tly b

y op

erat

ions

ne

eds.

Acc

ess

and

mai

nten

ance

act

iviti

es a

re fe

asib

le w

ith a

ll op

tions

but

wou

ld b

e sl

ight

ly e

asie

r with

a p

artia

lly e

mbe

dded

con

figur

atio

n w

ith re

duce

d ac

cess

and

egr

ess

trave

l dis

tanc

es fr

om p

oint

of e

ntry

to th

e bu

ildin

g.•

A ta

ble

is p

rovi

ded

in th

e re

port

show

ing

rela

tive

loca

tion

requ

irem

ents

of m

ajor

sys

tem

s

Slid

e9

NG

NP

TE

AMIn

itial

NG

NP

Layo

ut (P

artia

l Em

bedm

ent)

Slid

e10

NG

NP

TE

AMH

eat D

issi

patio

n to

Env

ironm

ent

from

the

Rea

ctor

•N

orm

al H

eat T

rans

fer t

o th

e E

nviro

nmen

t is

thro

ugh

RC

CS

and

P

HTS

/SH

TS, P

CS

, and

HP

S.

•D

BE

Hea

t tra

nsfe

r fro

m th

e re

acto

r is

thro

ugh

RC

CS

. Thi

s sc

enar

iodo

es n

ot re

ly o

n he

at tr

ansf

er th

roug

h th

e bu

ildin

g w

alls

.•

In th

e sp

ecifi

c B

DB

E c

ase

whe

re R

CC

S is

als

o lo

st, t

he re

acto

r de

cay

heat

is a

bsor

bed

in th

e R

eact

or C

itade

l wal

l.•

Sco

ping

cal

cula

tions

sho

w th

at w

alls

hav

e so

muc

h th

erm

al

capa

city

that

the

reac

tor a

ctua

lly s

tarts

to c

ool d

own

befo

re th

ete

mpe

ratu

re o

f the

out

side

of c

itade

l wal

l ris

es s

igni

fican

tly.

•P

BM

R h

as p

erfo

rmed

a s

impl

ified

ana

lysi

s to

dem

onst

rate

that

th

ere

is v

ery

little

diff

eren

ce b

etw

een

heat

tran

sfer

to S

oils

(Cla

y,S

and)

and

to

Air

out

side

the

Cita

del w

alls

.•

Gen

eral

find

ing

is th

at e

mbe

dmen

t dec

isio

n is

not

driv

en

sign

ifica

ntly

by

conc

ern

for h

eat d

issi

patio

n.

Slid

e11

NG

NP

TE

AMA

vera

ge c

oncr

ete

tem

pera

ture

at h

eigh

t co

rres

pond

ing

to m

iddl

e of

peb

ble

bed

RMY5

Slid

e 1

1

RM

Y5

Pete

r,O

ur m

odel

s w

ere

adju

sted

and

re-

run

as p

art

of t

he r

evie

w p

roce

ss a

nd t

he r

esul

ts d

iffer

slig

htly

, but

not

the

out

com

e.

I ha

ve in

sert

ed

the

new

gra

ph h

ere

for

you.

The

plot

"40

" is

a c

ompa

rativ

e pl

ot f

or t

he c

ase

whe

re t

he o

utsi

de t

empe

ratu

re o

f th

e ci

tade

l is

held

fix

ed a

t 40

deg

CRog

er M

You

ng, 9

/2/2

008

Slid

e12

NG

NP

TE

AM

Wat

er T

able

Effe

cts

•Th

e w

ater

tabl

e at

mos

t site

s w

ill lik

ely

be e

ncou

nter

ed a

t dep

ths

less

than

abo

ut 1

5m, e

xclu

ding

arid

regi

ons

•Fo

r site

s w

here

the

wat

er ta

ble

surfa

ce is

abo

ve th

e N

HS

B

foun

datio

n m

at th

e fo

llow

ing

issu

es to

be

addr

esse

d:–

Dew

ater

exc

avat

ion.

For

ver

y pe

rmea

ble

soil

or ro

ck m

ay n

eed

addi

tiona

l mea

sure

s to

con

trol

gro

undw

ater

inflo

w–

Obt

ain

grou

nd w

ater

with

draw

al p

erm

its–

Incr

ease

s so

il liq

uefa

ctio

n ris

k –

Des

ign

stru

ctur

es to

resi

st u

plift

pre

ssur

e (b

uoya

ncy)

and

sta

ticw

ater

pre

ssur

e–

Wat

erpr

oof t

he fo

unda

tion

mat

and

ext

erio

r wal

ls o

f the

NH

SB•

Gen

eral

find

ing

favo

rs m

inim

al e

mbe

dmen

t or p

artia

l em

bedm

ent o

nsi

tes

with

hig

h w

ater

tabl

e.

Slid

e13

NG

NP

TE

AM

Geo

tech

nica

l Con

stra

ints

and

Fo

unda

tion

Perf

orm

ance

•D

eep

embe

dmen

t req

uire

s ge

nera

lly m

ore

soph

istic

ated

exc

avat

ion

and

foun

datio

n co

nstru

ctio

n te

chni

ques

.•

On

Roc

k S

ites

cont

rolle

d bl

astin

g is

requ

ired

with

anc

hor b

oltin

gre

quire

d at

gre

ater

dep

ths

•O

n so

il si

tes

slop

ed o

pen

cut i

s fe

asib

le fo

r sha

llow

exc

avat

ions

.•

On

deep

soi

l site

s sh

eet p

iling

and

othe

r can

tilev

ered

type

s ar

efe

asib

le fo

r sha

llow

exc

avat

ions

. Bra

ced

wal

l is

requ

ired

at g

reat

erde

pths

.•

Eng

inee

red

back

fillin

g w

ith le

an c

oncr

ete

or c

ompa

cted

fill

incr

ease

sw

ith d

epth

.•

Res

ista

nce

to o

vertu

rnin

g im

prov

es w

ith d

epth

but

is n

egat

ed b

y in

crea

se in

buo

yanc

y ef

fect

at g

reat

er d

epth

s.•

Gen

eral

find

ing

favo

rs p

artia

l em

bedm

ent.

Slid

e14

NG

NP

TE

AM

Con

stru

ctio

n C

onsi

dera

tions

•D

ista

nce

belo

w G

rade

will

dire

ctly

impa

ct th

e le

vel o

f con

stru

ctio

nco

mpl

exity

•To

em

bed

a st

ruct

ure

on a

site

whe

re s

olid

rock

is a

t or n

ear t

hesu

rfac

e w

ould

sta

rt w

ith d

rillin

g an

d bl

astin

g an

d ro

ck re

mov

al to

the

dept

h of

the

botto

m o

f the

foun

datio

n. T

his

is a

cos

tly a

nd d

ange

rous

proc

ess

•Pu

mpi

ng s

yste

ms

are

still

requ

ired

to re

mov

e w

ater

that

may

see

pin

thro

ugh

rock

cra

cks

or th

at is

dep

osite

d by

rain

s •

On

soil

site

s, g

roun

d w

ater

and

sta

bilit

y of

the

soils

is o

f con

cern

.G

roun

d w

ater

mus

t be

pum

ped

away

from

the

exca

vatio

n. C

ontin

uous

pum

ping

is w

ell p

rove

n. D

ispo

sal o

f the

wat

er m

ust b

e co

nsid

ered

.Pr

otec

tion

from

cav

e-in

is re

quire

d. S

ide

wal

ls m

ay re

quire

ben

chin

g

Slid

e15

NG

NP

TE

AMCon

stru

ctio

n C

onsi

dera

tions

(con

t.)

•Sh

eet p

iling

to 1

2 m

or o

ther

eng

inee

red

deep

exc

avat

ion

supp

ort

tech

niqu

es b

eyon

d 12

m m

ay b

e re

quire

d.•

Dee

p em

bedm

ent w

ill im

pact

abi

lity

to u

se m

odul

ar c

onst

ruct

ion.

M

odul

es o

n lo

wer

leve

l wou

ld b

e re

quire

d to

be

com

plet

e fo

rear

lyin

stal

latio

n.•

Mod

ules

and

thei

r com

pone

nts

mus

t be

prot

ecte

d fr

om s

igni

fican

t ha

zard

s in

clud

ing

wea

ther

and

con

stru

ctio

n op

erat

ions

abo

ve

incl

udin

g co

ncre

te p

lace

men

t sin

ce th

ese

mod

ules

wou

ld h

ave

to b

epl

aced

in th

e st

ruct

ure

befo

re c

oncr

ete

floor

s ar

e in

stal

led.

Slid

e16

NG

NP

TE

AMCon

stru

ctio

n C

onsi

dera

tions

(con

t.)

•W

orki

ng b

elow

gra

de re

quire

s si

gnifi

cant

ven

tilat

ion

and

cont

inuo

usai

r mon

itorin

g to

ass

ure

wor

ker s

afet

y. S

ome

wel

ding

pro

cess

es

requ

ire th

e us

e of

iner

t gas

ses

such

as

argo

n fo

r a s

hiel

ding

med

ium

.A

rgon

gas

is e

xtre

mel

y da

nger

ous

in c

onfin

ed s

pace

s si

nce

it di

spla

ces

oxyg

en. A

rgon

is a

lso

heav

ier t

han

air a

nd te

nds

to fi

nd it

s w

ay to

the

low

est s

pace

s in

the

build

ing

volu

me

thus

pre

sent

ing

apo

tent

ial h

azar

d to

wor

kers

in th

ose

area

s. T

his

prob

lem

is w

orst

in a

fu

lly e

mbe

dded

cas

e w

here

con

stru

ctio

n op

enin

gs w

ould

not

be

able

to p

rom

ote

vent

ilatio

n.•

In s

umm

ary,

con

stru

ctio

n co

mpl

exity

and

cos

t in

crea

ses

expo

nent

ially

with

dep

th o

f em

bedm

ent

•G

ener

al fi

ndin

g si

gnifi

cant

ly fa

vors

min

imal

em

bedm

ent o

ver p

artia

lan

d fu

ll em

bedm

ent

Slid

e17

NG

NP

TE

AM

Cos

t Ben

efit

Con

side

ratio

n

•In

gen

eral

it e

xpec

ted

the

geot

echn

ical

and

foun

datio

n co

st w

illin

crea

seex

pone

ntia

lly w

ith d

epth

of e

mbe

dmen

t•

Ther

e is

a tr

ade

off b

etw

een

heav

y w

all t

hick

ness

es re

quire

d fo

rsec

urity

and

resi

stan

ce fo

r na

tura

l phe

nom

ena

abov

e gr

ade

com

pare

d w

ith

incr

ease

d w

all t

hick

ness

requ

ired

to re

sist

soi

l for

ces

as e

mbe

dmen

tin

crea

ses.

Thi

s w

ill be

revi

ewed

in m

ore

deta

il in

the

repo

rt.•

A co

mpa

rison

has

bee

n m

ade

of re

lativ

e co

st o

f the

pro

pose

d em

bedm

ent

optio

ns fo

r rec

tang

ular

and

circ

ular

stru

ctur

es a

nd tw

o si

te o

ptio

ns, (

Roc

k w

ith lo

w w

ater

tabl

e, a

nd D

eep

soil

with

hig

h w

ater

tabl

e).

•K

ey C

ost d

iffer

entia

l ele

men

ts a

re e

xcav

atio

n, b

ackf

ill, d

ewat

erin

g,su

bsur

face

foun

datio

n a

nd e

xter

ior c

oncr

ete

wal

ls, a

bove

and

bel

owgr

ound

.•

Initi

al fi

ndin

g is

that

Ful

l em

bedm

ent i

s si

gnifi

cant

ly m

ore

cost

ly.

•D

eep

exca

vatio

n on

a R

ock

site

is m

ore

cost

ly th

an o

n a

soil

site

.

Slid

e18

NG

NP

TE

AM

REA

CTO

R B

UIL

DIN

G –

OPE

N C

UT

EXC

AVA

TIO

NEX

CA

VATI

ON

VOLU

ME

vs. E

MB

EDD

ED D

EPTH

0

2000

00

4000

00

6000

00

8000

00

1000

000

1200

000

1400

000

1600

000

1800

000

2000

000

010

2030

4050

6070

Dep

th (M

)

Volume (M3)

Serie

s1

Slid

e19

NG

NP

TE

AM

NO

RM

ALI

ZED

REL

ATI

VE C

OST

R

EAC

TOR

BU

ILD

ING

EM

BED

MEN

T

NO

RM

ALIZ

ED R

ELAT

IVE

CO

ST O

F EM

BED

MEN

T

0.00

5.00

10.0

0

15.0

0

20.0

0

25.0

0

010

2030

4050

6070

Embe

dmen

tDep

th (M

)

Base Cost Factor

Circ

ular

Ext

erio

r Wal

l, R

ock

belo

w G

rade

Circ

ular

Ext

erio

r Wal

l, So

il bel

ow

Gra

de

Squ

are

Ext

erio

r Wal

l, R

ock

belo

w G

rade

Squ

are

Ext

erio

r Wal

l, So

il bel

ow

Gra

de

Slid

e20

NG

NP

TE

AM

ESTI

MA

TED

CA

PITA

L C

OST

-K

EY E

LEM

ENTS

REA

CTO

R B

UIL

DIN

G E

MB

EDM

ENT

ESTI

MAT

ED C

APITA

L CO

ST -

KEY

ELEM

ENTS

$0

$100

,000

$200

,000

$300

,000

$400

,000

$500

,000

$600

,000

010

2030

4050

6070

Embe

dmen

tDep

th (M

)

Dollars (in 000's)

Circ

ular E

xterio

r Wall

, Roc

k bel

ow G

rade

Circ

ular E

xterio

r Wall

, Soi

l belo

w Gr

ade

Squa

re E

xterio

r Wall

, Roc

k bel

ow G

rade

Squa

re E

xterio

r Wall

, Soi

l belo

w Gr

ade

Slid

e21

NG

NP

TE

AM

Des

ign

Bas

is T

hrea

t and

Mal

evol

ent H

azar

ds

•Fo

llow

ing

Sept

11,

200

1 N

RC

issu

ed IC

Ms

to b

ette

r def

ine

secu

rity

and

prot

ectiv

e re

quire

men

ts. T

hese

ICM

sar

e ge

nera

lly s

afeg

uard

s in

form

atio

n.•

Lice

nsee

s ar

e re

ques

ted

to a

sses

s is

sues

suc

h as

larg

e ai

rpla

necr

ash,

larg

e fir

es, a

nd e

xplo

sion

s as

bey

ond

desi

gn b

asis

eve

nts.

•A

sses

smen

t gui

delin

es a

re in

clud

ed in

the

“Met

hodo

logy

for

Perf

orm

ing

Airc

raft

Impa

ct A

sses

smen

ts fo

r “N

ew P

lant

Des

igns

”so

on to

be

issu

ed a

s N

EI 0

8-13

•If

the

plan

t is

tota

lly u

nder

grou

nd, i

t pre

sent

s no

targ

et fo

r airc

raft

cras

h an

d no

furt

her a

sses

smen

t is

need

ed o

ther

than

for c

rash

in

duce

d vi

brat

ion

effe

cts

on in

-pla

nt e

quip

men

t. •

If po

rtio

ns a

re a

bove

gra

de, t

hen

a be

yond

des

ign

basi

s as

sess

men

tm

ust b

e m

ade

to s

how

no

sign

ifica

nt ra

dion

uclid

e re

leas

e (r

eact

orco

re, n

ew fu

el, a

nd s

pent

fuel

han

dlin

g an

d st

orag

e sy

stem

s)

Slid

e22

NG

NP

TE

AM

Des

ign

Bas

is T

hrea

t and

Mal

evol

ent H

azar

ds

(con

t.)•

Rea

ctor

bui

ldin

gs in

corp

orat

ing

robu

st s

truc

tura

l des

ign

for s

eism

ican

d D

BT

resi

stan

ce c

an g

ener

ally

sho

w a

ccep

tabl

e ai

rcra

ft cr

ash

asse

ssm

ent r

esul

ts.

•Ev

ent m

itiga

tion

feat

ures

and

lice

nsee

str

ateg

ies

mus

t be

prov

ided

to

NR

C d

urin

g th

e C

ombi

ned

Lice

nse

appl

icat

ion

phas

e of

pla

nt

depl

oym

ent i

n ac

cord

ance

with

pro

pose

d R

ule

10C

FR50

.54(

hh)2

•

Larg

e fir

es a

re d

efin

ed a

s th

ose

grea

ter t

han

thos

e de

fined

in ty

pica

lFH

A.

They

affe

ct la

rge

port

ions

of t

he p

lant

, wel

l in

exce

ss o

ftho

seha

ving

thei

r orig

in fr

om o

n-si

te s

ourc

es.

No

expl

icit

initi

atio

n m

echa

nism

is a

ssum

ed fo

r the

se fi

res.

His

toric

ally

, one

sou

rce

ofth

ese

larg

e fir

es is

the

rele

ase

of je

t fue

l fro

m a

larg

e co

mm

erci

alai

rcra

ft cr

ash.

Slid

e23

NG

NP

TE

AM

Des

ign

Bas

is T

hrea

t and

Mal

evol

ent H

azar

ds

(con

t.)•

Rul

es o

f thu

mb:

–M

inim

ize

or e

limin

ate

the

effe

ctiv

e ta

rget

are

a fo

r airc

raft

cras

h–

In a

dditi

on to

gen

eral

str

uctu

ral r

obus

tnes

s of

the

reac

tor b

uild

ing

for s

eism

ic, e

nhan

ce th

e pe

netr

atio

n re

sist

ance

of r

eact

or b

uild

ing

stru

ctur

es.

–M

axim

ize

the

inhe

rent

and

pas

sive

fuel

coo

ling

and

radi

atio

n co

ntai

nmen

t fea

ture

s of

the

plan

t.–

Max

imiz

e se

para

tion

of re

dund

ant s

afet

y fu

nctio

n in

itiat

ion

feat

ures

.–

Max

imiz

e se

para

tion

of s

afet

y an

d de

fens

e in

dep

th s

yste

m

feat

ures

.–

Ensu

re th

at a

ltern

ate

sour

ces

and

mea

ns o

f fue

l coo

ling

are

avai

labl

e fo

llow

ing

a be

yond

des

ign

basi

s th

reat

eve

nt.

•In

itial

find

ing

favo

rs fu

ll em

bedm

ent.

But

it is

foun

d th

at p

rote

ctio

n fo

r se

ism

ic a

nd p

rote

ctio

n fr

om n

atur

al h

azar

ds w

ould

mos

t lik

ely

prov

ide

adeq

uate

resi

stan

ce.

Slid

e24

NG

NP

TE

AM

Nat

ural

Phe

nom

enon

1.W

ind

Load

ing

will

follo

w A

SC

E 7

and

Inte

rnat

iona

l Bui

ldin

g C

ode.

Hur

rican

e w

inds

of 6

5 m

/s( 3

sec

ond

gust

s) a

t 10

m a

bove

gra

de w

ith

impo

rtanc

e fa

ctor

s as

sign

ed b

ased

on

build

ing

func

tion.

2.

Rec

ent R

.G. 1

.76

Torn

ado

Load

ing

Crit

eria

are

less

sev

ere

for p

ress

ure

drop

and

win

d sp

eed

than

UR

D.

Use

R.G

. 1.7

6 fo

r the

se c

ondi

tions

.

EPR

I-UR

DR

G-1

.76

•M

ax T

orna

do W

ind

Spe

ed13

4m

/s10

3m

/s•

Max

Rot

atio

nal S

peed

107

m/s

82m

/s•

Max

Tra

nsla

tiona

l Spe

ed27

m/s

21m

/s•

Rad

ius

of M

ax R

otat

iona

l Spe

ed45

.7 m

45.7

m•

Max

imum

Pre

ssur

e D

rop

138

mb

83m

b•

Rat

e of

Pre

ssur

e D

rop

83m

b/se

c37

mb/

sec

Slid

e25

NG

NP

TE

AM

Nat

ural

Phe

nom

enon

•To

rnad

o M

issi

les

per R

.G. 1

.76

Mas

sV

eloc

ity–

Sch

40 P

ipe

x 4.

57m

long

13

0 kg

41

m/s

–A

utom

obile

1810

kg

41m

/s–

Solid

sph

ere

0.06

69 k

g8

m/s

Not

es:

1.Ve

loci

ty li

sted

are

in h

oriz

onta

l dire

ctio

n. V

ertic

al v

eloc

ities

sha

ll be

67

perc

ent o

f the

ho

rizon

tal v

eloc

ities

.2.

Aut

omob

ile t

o a

heig

ht o

f 9.1

4 m

onl

y.3.

Reg

ion

I mis

sile

s us

ed.

4.R

einf

orce

d co

ncre

te w

alls

abo

ut 0

.75

m th

ick

are

typi

cally

ade

quat

e fo

r tor

nado

m

issi

les.

•Fl

ood

and

Dam

Bre

ak –

Site

to b

e ch

osen

with

gra

de to

be

0.3

m a

bove

max

imum

P

roba

ble

Floo

d in

clud

ing

dam

bre

ak .

•S

eism

ic re

spon

se s

pect

ra to

be

dete

rmin

ed b

ased

on

Sei

smic

Pro

babi

listic

Haz

ard

Ass

ess.

The

UR

D p

ropo

ses

a ge

neric

SS

E c

ompr

isin

g a

sing

le g

roun

d m

otio

n sp

ectru

m c

onfo

rmin

g to

a p

eak

grou

nd a

ccel

erat

ion

of 0

.3 g

.

•Fi

ndin

g fa

vors

em

bedm

ent,

but i

t is

com

mon

to d

esig

n of

stru

ctur

es to

resi

st th

ese

haza

rds.

Slid

e26

NG

NP

TE

AM

Issu

es R

elat

ive

to S

eism

ic A

naly

sis

& D

esig

n•

Seis

mic

Des

ign

perf

orm

ed in

acc

orda

nce

with

EPR

I-UR

D•

Seis

mic

Cla

ssifi

catio

n Sy

stem

con

sist

ent w

ith N

ucle

ar R

egul

ator

yG

uide

1.2

9.•

Seis

mic

Cat

egor

y–

Seis

mic

Cat

egor

y I –

This

cla

ssifi

catio

n in

clud

es a

ll st

ruct

ures

, sy

stem

s an

d co

mpo

nent

s w

hose

saf

ety

clas

s is

SC

-1, S

C-2

, or S

C-

3. S

eism

ic c

ateg

ory

I sha

ll al

so in

clud

e sp

ent f

uel s

tora

ge p

ool

stru

ctur

es in

clud

ing

all f

uel r

acks

.–

Seis

mic

Cat

egor

y II

–Th

is c

lass

ifica

tion

appl

ies

to a

ll pl

ant

stru

ctur

es, s

yste

ms

and

com

pone

nts

whi

ch p

erfo

rm n

o nu

clea

r sa

fety

func

tion

but w

hose

failu

re c

ould

deg

rade

SC

-1, S

C-2

, and

SC

-3 s

truc

ture

s, s

yste

ms

and/

or c

ompo

nent

s.–

Non

-sei

smic

–Th

is c

lass

ifica

tion

incl

udes

all

stru

ctur

es th

at d

o no

t fal

l int

o Se

ism

ic C

ateg

ory

I or S

eism

ic C

ateg

ory

II st

ruct

ures

.•

No

OB

E Lo

ad C

ase

•SS

E co

mpr

isin

g a

sing

le g

roun

d m

otio

n sp

ectr

um c

onfo

rmin

g to

R

egul

ator

y G

uide

1.6

0 an

chor

ed to

a 0

.3g

peak

gro

und

acce

lera

tion.

This

pea

k gr

ound

acc

eler

atio

n w

ill a

llow

the

NG

NP

to b

e co

nstr

ucte

d at

m

ost s

ites

in th

e co

ntin

enta

l Uni

ted

Stat

es e

ast o

f the

Roc

ky M

ount

ains

.•

The

desi

gn re

spon

se s

pect

ra s

hall

be in

acc

orda

nce

with

Reg

ulat

ory

Gui

de 1

.60

–“D

esig

n R

espo

nse

Spec

tra

for S

eism

ic D

esig

n of

Nuc

lear

Po

wer

Pla

nts”

with

a ti

me

hist

ory

to e

nvel

op th

e de

sign

spe

ctra

.

Slid

e27

NG

NP

TE

AM

Issu

es R

elat

ive

to S

eism

ic A

naly

sis

& D

esig

n

Ana

lysi

s &

Des

ign

•D

esig

n of

Sei

smic

Cat

egor

y I s

truc

ture

s sh

all f

ollo

w th

e gr

ound

mot

ion

char

acte

ristic

s sp

ecifi

ed p

revi

ousl

y, th

e an

alyt

ical

tech

niqu

es s

peci

fied

in

ASC

E St

anda

rd 4

-98

and

mee

t all

qual

ity a

ssur

ance

requ

irem

ents

of 1

0 C

FR,

App

endi

x B

. D

esig

n of

Sei

smic

Cat

egor

y I c

oncr

ete

stru

ctur

es s

hall

be in

ac

cord

ance

with

AC

I 349

-06.

Des

ign

of S

eism

ic C

ateg

ory

I ste

el s

truc

ture

s sh

all b

e in

acc

orda

nce

with

AIS

C N

690L

-03.

•D

esig

n of

sei

smic

cat

egor

y II

stru

ctur

es s

hall

utili

ze th

e sa

me

anal

ysis

and

de

sign

cod

es a

s th

ose

used

for s

eism

ic c

ateg

ory

I str

uctu

res;

how

ever

, the

se

stru

ctur

es m

ay b

e co

nstr

ucte

d to

the

requ

irem

ents

of t

he n

on-n

ucle

ar b

uild

ing

code

s. I

n ge

nera

l, th

ese

requ

irem

ents

will

invo

ke th

e In

tern

atio

nal B

uild

ing

Cod

e (IB

C-0

6) o

r the

ver

sion

spe

cifie

d by

the

loca

l bui

ldin

g au

thor

ity.

•D

esig

n of

Sei

smic

Cat

egor

y III

str

uctu

res

shal

l fol

low

the

anal

ytic

alre

quire

men

ts o

f IB

C-0

6. I

BC

-06

spec

ifies

that

AC

I-318

-05

by u

sed

for t

he

desi

gn o

f con

cret

e st

ruct

ures

and

that

AIS

C–

ASD

/LR

FD S

teel

Con

stru

ctio

n m

anua

l, 13

th E

ditio

n be

use

d fo

r ste

el d

esig

n.

Slid

e28

NG

NP

TE

AM

Estim

ated

Inpu

t Bas

e A

ccel

erat

ion

of R

eact

or

Bui

ldin

g D

ue to

Em

bedm

ent D

epth

0.3

g G

roun

d Su

rfac

e A

ccel

erat

ion

-SO

IL S

ITE

RE

AC

TO

R B

UIL

DIN

G

EM

BE

DM

EN

TPE

RC

EN

TA

NG

E

EST

IMA

TE

D IN

PUT

A

CC

EL

ER

AT

ION

AT

BA

SE

OF

RE

AC

TO

R B

UIL

DIN

G

(g’s

)

EST

IMA

TE

DPE

AK

AC

CE

LE

RA

TIO

NT

OP

OF

RE

AC

TO

RB

UIL

DIN

G0

0.30

01.

0 (A

SSU

ME

D)

100.

276

0.92

200.

252

0.84

300.

228

0.76

400.

204

0.68

500.

180

0.60

100

0.18

00.

60

Slid

e29

NG

NP

TE

AM

Estim

ated

Inpu

t Bas

e A

ccel

erat

ion

of R

eact

or

Bui

ldin

g D

ue to

Em

bedm

ent D

epth

0.3

g G

roun

d Su

rfac

e A

ccel

erat

ion

-SO

IL S

ITE

Est

imat

ed P

eak

Acc

eler

atio

nTo

p of

RB

0.8

4 g

Est

imat

ed P

eak

Acc

eler

atio

nR

eact

orTo

p of

RB

0.6

0 g

Build

ing

(RB

)h

Rea

ctor

hBu

ildin

gG

rade

(RB

)

20 %

h E

mbe

dmen

t

Est

imat

ed B

ase

Inpu

t50

% h

Acc

eler

atio

n 0.

252

gE

mbe

dmen

t

Est

imat

ed B

ase

Inpu

tA

ccel

erat

ion

0.18

0 g

Slid

e30

NG

NP

TE

AM

Nat

ural

Geo

logi

cal P

heno

men

a

•Se

ism

ic H

azar

ds–

Ear

thqu

ake

haza

rd m

ajor

con

side

ratio

n to

site

sel

ectio

n–

Em

bedd

ed m

at fo

unda

tions

bea

ring

on th

ick

soil

are

subj

ect t

o so

il am

plifi

catio

n.–

Mot

ion

at th

e m

at is

gre

ater

than

mot

ion

at to

p of

bed

rock

.–

Site

gro

und

mot

ion

esta

blis

hed

usin

g N

RC

met

hods

in 1

0 C

FR 1

00.2

3“G

eolo

gic

& S

eism

ic S

iting

Crit

eria

”–

Pro

babi

listic

met

hods

for s

ite s

eism

ic h

azar

d in

Reg

ulat

ory

Gui

de 1

.208

–

Stru

ctur

al d

esig

n in

hig

h se

ism

ic a

reas

(e.g

. par

ts o

f the

wes

tern

U.S

. and

in

prox

imity

to th

e N

ew M

adrid

Zon

e in

the

Mis

siss

ippi

Val

ley)

requ

ire m

ore

robu

st

seis

mic

rest

rain

ts w

heth

er fo

unde

d on

rock

or s

oil

–R

ock

site

s ar

e no

t sub

ject

to a

mpl

ifica

tion

effe

cts

–Li

quef

actio

n of

loos

e sa

tura

ted

soil

can

be tr

igge

red

by e

arth

quak

e–

Roc

k si

tes

are

not s

ubje

ct to

liqu

efac

tion

Slid

e31

NG

NP

TE

AM

Nat

ural

Geo

logi

cal P

heno

men

a (c

ont)

•G

eolo

gic

Haz

ards

–A

void

site

loca

tions

that

may

be

subj

ect t

o:•

Cap

able

faul

ts a

nd fa

ults

exh

ibiti

ng s

urfa

ce ru

ptur

e•

Kar

stic

terra

in fo

rmed

by

diss

olut

ion

of s

olub

le ro

cks/

soil

such

as

limes

tone

, sal

t and

gyp

sum

•Su

rface

col

laps

e re

sulti

ng fr

om u

nder

grou

nd m

inin

g op

erat

ions

•

Act

ive

or fo

rmer

slo

pe in

stab

ility

•La

ndsl

ide

gene

rate

d flo

od w

aves

or s

eich

eflo

od w

aves

•A

ctiv

e or

rece

nt v

olca

nism

•

Sub

side

nce

resu

lting

from

the

extra

ctio

n of

gro

undw

ater

or

hydr

ocar

bons

•Fu

ture

sub

surfa

ce m

inin

g•

Hig

h st

ress

con

cent

ratio

ns in

rock

•W

eak

and

very

com

pres

sibl

e so

il

Slid

e32

NG

NP

TE

AM

Soil

Stru

ctur

e In

tera

ctio

n

•A

n ad

ditio

nal c

onsi

dera

tion

of th

is s

tudy

is to

eva

luat

e th

e ef

fect

s of

stru

ctur

e em

bedm

ent o

n th

e se

ism

ic in

put g

roun

d m

otio

n at

the

base

of t

he s

truct

ure.

In

gene

ral,

embe

ddin

g th

e st

ruct

ure

redu

ces

the

inpu

t mot

ion

at th

eba

se o

f the

st

ruct

ure;

ther

eby,

redu

cing

the

seis

mic

load

. H

owev

er, t

here

are

lim

its to

this

effe

ct.

Bas

ed o

n pu

blis

hed

liter

atur

e, it

is re

ason

ably

con

serv

ativ

e to

ass

ume

that

em

bedm

ent w

ill n

ot re

duce

the

inpu

t bas

e gr

ound

mot

ion

to le

ss th

an a

ppro

xim

atel

y 60

% o

f the

sur

face

gro

und

mot

ion.

For

the

assu

med

reac

tor b

uild

ing,

this

lim

it is

re

ache

d at

a d

epth

of a

ppro

xim

atel

y 30

m.

In a

dditi

on, A

SC

E 4

sta

tes

that

the

top

6 m

or o

ne-h

alf o

f the

em

bedm

ent d

epth

, whi

chev

er is

less

, mus

t be

negl

ecte

d in

the

soil

stru

ctur

e in

tera

ctio

n fo

rmul

atio

n. B

ased

on

the

abov

e, it

is e

stim

ated

that

littl

e or

no

redu

ctio

n w

ill b

e re

aliz

ed a

t em

bedm

ents

exce

edin

g 36

m o

r abo

ut 5

0%

embe

dmen

t of t

he re

acto

r bui

ldin

g.

•It

is n

ot u

nrea

listic

to a

ssum

e th

at th

e in

put m

otio

n at

the

base

of t

he s

truct

ure

varie

s lin

early

from

100

% a

t gra

de to

60%

at 3

6 m

. •

Ther

e ar

e ot

her f

acto

rs th

at c

ontri

bute

to a

mpl

ifica

tion

of th

e pe

ak g

roun

d ac

cele

ratio

n. A

s an

exa

mpl

e, th

e st

iffne

ss a

nd la

yer t

hick

ness

of t

he s

oil s

igni

fican

t im

pact

the

ampl

ifica

tion

fact

or.

Slid

e33

NG

NP

TE

AMChe

mic

al R

elea

ses

and

Expl

osio

ns

•R

egul

ator

y G

uide

1.7

8 re

quire

s co

nsid

erat

ion

of q

uant

ities

gre

ater

than

100

pou

nds

(45k

g) o

f haz

ardo

us c

hem

ical

s w

ithin

0.3

mile

(4

83m

) of t

he c

ontro

l roo

m•

Sta

tiona

ry a

nd M

obile

Sou

rces

will

be c

onsi

dere

d•

Dis

pers

ion

mod

elin

g w

ill b

e us

ed to

est

ablis

h pl

ume

conc

entra

tions

vs. d

ista

nce

from

Sou

rce.

•C

orro

sive

che

mic

als,

Tox

ic g

ases

, Asp

hyxi

ates

, Fla

mm

able

gas

es,

Oxy

gen

rele

ases

will

be c

onsi

dere

d.•

NU

RE

G/C

R-6

944

“PIR

Ts”w

ill b

e co

nsid

ered

. •

Con

cern

for h

eavi

er th

an a

ir ga

s re

leas

es e

ncro

achi

ng o

n re

acto

rbu

ildin

g.•

Con

cern

for c

once

ntra

tion

of fl

amm

able

gas

es in

low

lyin

g lo

catio

ns

Slid

e34

NG

NP

TE

AMChe

mic

al R

elea

ses

and

Expl

osio

ns

•C

once

rn fo

r fire

s an

d de

flagr

atio

n ev

ent n

ear t

he R

eact

or b

uild

ing

site

.•

Con

cern

for c

loud

s of

exp

losi

ve g

ases

col

lect

ing

and

deto

natin

g in

prox

imity

to re

acto

r bui

ldin

g.•

Pro

paga

tion

of h

azar

dous

che

mic

al th

roug

h se

cond

ary

or te

rtiar

y sy

stem

s to

the

reac

tor b

uild

ing

•E

xplo

sion

Ana

lysi

s pe

r R.G

. 1.9

1 re

quire

d to

est

ablis

h sa

fe d

ista

nce

and

quan

titie

s•

Initi

al F

indi

ng is

that

em

bedd

ed b

uild

ing

coul

d be

slig

htly

mor

evu

lner

able

to c

olle

ctio

n of

hea

vier

than

air

gase

s. It

is e

xpec

ted

that

fe

atur

es re

quire

d fo

r oth

er it

ems

will

prov

ide

prot

ectio

n.

Slid

e35

NG

NP

TE

AM

Sum

mar

y of

Fin

ding

s

Maj

or c

ost D

river

Min

imal

or p

artia

l em

bedm

ent p

refe

rred

C

onst

ruct

ion

Con

side

ratio

ns

Maj

or c

ost d

river

Par

tial e

mbe

dmen

t pr

efer

red

to b

alan

ce

resi

stan

ce to

ove

rturn

ing

Geo

tech

nica

l Con

stra

ints

an

d Fo

unda

tion

Per

form

ance

Mod

erat

eM

inim

al e

mbe

dmen

t pr

efer

red

on s

oil s

ites

with

hi

gh w

ater

tabl

e

Wat

er T

able

Effe

cts

Min

orD

iffer

ence

in h

eat t

rans

fer

not s

igni

fican

tH

eat D

issi

patio

n to

E

nviro

nmen

t

Min

or to

Mod

erat

eP

artia

l em

bedm

ent

pref

erre

d (b

ut n

ot re

quire

d)to

allo

w q

uick

acc

ess

and

egre

ss

Ope

ratio

nal N

eeds

(Saf

ety

equi

pmen

t lay

out)

Rel

ativ

e Im

port

ance

to

Des

ign

Sele

ctio

nIn

itial

Fin

ding

Tech

nica

l Iss

ue

Slid

e36

NG

NP

TE

AM

Sum

mar

y of

Fin

ding

s Mod

erat

eFa

vors

Min

imal

em

bedm

ent

due

to c

once

rn fo

r int

rusi

on

of h

eavi

er th

an a

ir ga

ses

Che

mic

alR

elea

ses

and

Exp

losi

ons

Mod

erat

e im

porta

nt tr

ade

off

Favo

rs E

mbe

dmen

t but

ex

pect

less

adv

anta

ge a

t gr

eate

r dep

ths

Nat

ural

Geo

logi

cal

Phe

nom

ena

Mod

erat

eFa

vors

Em

bedm

ent t

o re

duce

haz

ard

prot

ectio

n bu

t ot

her a

ltern

ativ

es a

re

feas

ible

.

Nat

ural

Phe

nom

ena

Maj

orFa

vors

Em

bedm

ent b

ut o

ther

al

tern

ativ

es a

re fe

asib

le.

Des

ign

Bas

is T

hrea

t and

M

alev

olen

tHaz

ards

Maj

or c

onsi

dera

tion

Cle

arly

favo

rs le

ss

embe

dmen

tC

ost B

enef

it C

onsi

dera

tion

Rel

ativ

e Im

port

ance

to

Des

ign

Sele

ctio

nIn

itial

Fin

ding

Tech

nica

l Iss

ue

Slid

e37

NG

NP

TE

AM

Obj

ectiv

e R

anki

ngW

eigh

tO

bjec

tive

100

Tota

l Wei

ght P

erce

nt

10C

onst

ruct

ion

Con

side

ratio

ns (N

ote

10)

10G

eote

chni

cal C

onst

rain

ts a

nd F

ound

atio

n Pe

rform

ance

(N

ote

9)

10W

ater

Tab

le E

ffect

s (N

ote

8)

10C

ost B

enef

it C

onsi

dera

tion

(Not

e 7)

Cos

t and

Con

stru

ctio

n C

ompl

exity

5C

hem

ical

Rel

ease

s an

d Ex

plos

ions

(Not

e 6)

10N

atur

al G

eolo

gica

l Phe

nom

ena

(Not

e 5)

15N

atur

al P

heno

men

a (N

ote

4)

15D

esig

n Ba

sis

Thre

at a

nd M

alev

olen

t Haz

ards

(Not

e 3)

5H

eat D

issi

patio

n to

Env

ironm

ent (

Not

e 2)

Safe

ty, I

nves

tmen

t Pro

tect

ion,

and

Sec

urity

10O

pera

tiona

l Nee

ds (S

afet

y eq

uipm

ent l

ayou

t) (N

ote

1)

Des

ign

for O

pera

tions

and

Pro

per S

yste

m M

echa

nica

l Fun

ctio

ns

Slid

e38

NG

NP

TE

AM

Obj

ectiv

e R

anki

ng W

eigh

t Not

es

1)D

esig

n fo

r Ope

ratio

nal e

ffect

s (R

eact

or p

rote

ctio

n, a

cces

s fo

r ref

uelin

g, e

tc.)

for n

orm

al a

nd e

mer

genc

y op

erat

ions

,DB

Es,

and

mai

nten

ance

act

iviti

es. R

elat

ive

loca

tions

of s

ub-s

yste

ms

with

resp

ect t

o re

acto

r ar

e co

nsid

ered

. 2)

The

abilit

y of

the

build

ing

to tr

ansf

er h

eat t

o th

e en

viro

nmen

t eith

er to

gro

und

or to

am

bien

t air

are

asse

ssed

.3)

The

elem

ent a

ddre

sses

pro

tect

ion

agai

nst t

he d

esig

n ba

sis

safe

guar

ds a

nd s

ecur

ity th

reat

DB

T an

d th

e be

yond

des

ign

thre

at im

pose

d by

a p

oten

tial n

on a

ccid

enta

l airc

raft

strik

e, o

r oth

er B

DB

E.

4)N

atur

al P

heno

men

on H

azar

ds (N

PH

), su

ch a

s to

rnad

o (in

clud

ing

mis

sile

s) a

nd fl

oodi

ng a

re c

onsi

dere

d.5)

The

pote

ntia

l im

pact

s th

at n

atur

al g

eolo

gica

l haz

ards

suc

h as

dis

solu

tion

feat

ures

in s

olub

le ro

ck, w

eak

com

pres

sibl

e so

ils, s

lope

inst

abilit

y re

sulti

ng in

land

slid

e po

tent

ial,

and

lique

fact

ion

and

othe

r ear

thqu

ake

indu

ced

phen

omen

a ar

e co

nsid

ered

6)C

onsi

dera

tion

for t

he h

azar

ds o

f che

mic

als

stor

ed a

nd o

r tra

nspo

rted

in ra

nge

of th

e R

eact

or, t

he

pote

ntia

l effe

cts,

and

miti

gatin

g fe

atur

es th

at c

ould

be

utiliz

ed. T

his

type

of c

once

rn d

oes

ofte

n m

ater

ializ

e w

ith o

pera

ting

faci

litie

s lo

cate

d of

act

ive

indu

stria

l are

as. T

he p

oten

tial f

ro h

ydro

gen

expl

osio

ns is

add

ress

ed h

ere.

7)C

ost b

enef

it in

clud

ing

key

elem

ents

of r

elat

ive

capi

tal c

ost w

ill b

e de

velo

ped

for v

ario

us e

mbe

dmen

t op

tions

and

com

pare

d in

a c

ost b

enef

it ta

ble.

8)G

roun

dwat

er le

vels

are

eva

luat

ed w

ith re

gard

to im

pact

on

stru

ctur

al d

esig

n, c

onst

ruct

ion

dew

ater

ing

sche

mes

and

nee

d fo

r wat

erpr

oofin

g.9)

A ra

nge

of fo

unda

tion

mat

eria

ls is

con

side

red

with

rega

rd to

sta

tic a

nd d

ynam

ic p

rope

rties

and

fo

unda

tion

perfo

rman

ce.

10)

The

effe

cts

of th

e em

bedm

ent d

epth

on

cons

truct

abilit

y us

ing

mod

ern

tech

niqu

es s

uch

as m

odul

ariz

atio

n ar

e co

nsid

ered

. The

exc

avat

ion

met

hod

used

will

requ

ire a

n ev

alua

tion

of o

verb

urde

n th

ickn

ess,

el

evat

ion

of c

ompe

tent

rock

, site

spa

ce c

onst

rain

ts, a

nd c

ost a

nd s

ched

ule

impa

cts

Slid

e39

NG

NP

TE

AM

Obj

ectiv

e R

anki

ng S

umm

ary

Diff

icul

ties

with

wat

er

tabl

e an

d co

nstru

ctio

n38

5So

il Si

te F

ull E

mbe

dmen

t

Best

Sco

re

650

Soil

Site

Par

tial E

mbe

dmen

t

Bene

fit w

ith re

duce

d w

ater

tabl

e co

ncer

n59

0So

il Si

te M

inim

um E

mbe

dmen

t

Less

wat

er ta

ble

conc

ern;

Exc

avat

ion

Cos

t unf

avor

able

455

Roc

k Si

te F

ull E

mbe

dmen

t

Best

Sco

re

620

Roc

k Si

te P

artia

l Em

bedm

ent

Max

exp

osur

e to

ha

zard

s52

0R

ock

Site

Min

imum

Em

bedm

ent

Rem

arks

Tota

l Wei

ghte

d Sc

ore

Des

ign

Alte

rnat

ive

Slid

e40

NG

NP

TE

AM

Alte

rnat

ive

Site

Con

ditio

ns

Roc

k Si

te•

Site

is c

hara

cter

ized

by

soun

d ro

ck a

t or n

ear g

rade

(< 1

5 m

dee

p)•

Foun

datio

n m

at fo

r the

Nuc

lear

Hea

t Sou

rce

Bui

ldin

g (N

HS

B) w

ill be

on

soun

d ro

ck.

•Fu

ll em

bedm

ent o

f the

NH

SB

will

requ

ire a

dee

p ex

cava

tion

in ro

ck (4

5 –

60 m

)•

Min

imal

or p

artia

l em

bedm

ent m

ay b

e pr

efer

red

due

to:

low

er e

xcav

atio

n co

stre

duce

d ne

ed fo

r gro

undw

ater

con

trol (

dew

ater

ing

and

wat

erpr

oofin

g)re

duce

d st

orm

wat

er c

ontro

l iss

ues

durin

g co

nstru

ctio

n•

Sei

smic

des

ign

is s

impl

ified

(soi

l-stru

ctur

e in

tera

ctio

n an

alys

is n

ot re

quire

d) fo

r roc

k w

ith s

hear

wav

e ve

loci

ty >

2800

m/s

•Li

quef

actio

n is

not

an

issu

e fo

r NH

SB

but

, may

be

for a

t-gra

de s

truct

ures

on

soil

•R

egio

nal C

onsi

dera

tions

: R

ock

site

s m

ost l

ikel

y to

be

foun

d in

upl

and

area

s of

NE

, S

E, N

W a

nd S

Cen

tral U

S.

Slid

e41

NG

NP

TE

AMA

ltern

ativ

e Si

te C

ondi

tions

(con

t.)So

il ov

er S

hallo

w R

ock

•S

ite c

hara

cter

ized

by

soil

over

lyin

g so

und

rock

at 1

5 –

50 m

dep

th•

Foun

datio

n m

at fo

r the

NH

SB

on

soun

d ro

ck•

Full

embe

dmen

t will

requ

ire e

xcav

atio

n in

soi

l and

, pos

sibl

y, in

rock

•S

oil e

xcav

atio

n w

ill ha

ve m

ax s

ide

slop

es o

f 1.5

H:1

V o

r 2H

:1V

requ

iring

a m

uch

larg

er e

xcav

atio

n as

the

rock

foun

ding

laye

r bec

omes

dee

per (

or e

mbe

dmen

tgr

eate

r).

Exc

avat

ion

in ro

ck m

ay b

e ne

ar-v

ertic

al; b

ut, m

ay re

quire

eng

inee

red

supp

ort

•M

inim

al o

r par

tial e

mbe

dmen

t may

be

pref

erre

d du

e to

:–

low

er e

xcav

atio

n co

st–

redu

ced

need

for g

roun

dwat

er c

ontro

l (de

wat

erin

g an

d w

ater

proo

fing)

–re

duce

d st

orm

wat

er c

ontro

l iss

ues

durin

g co

nstru

ctio

n•

Sei

smic

des

ign

is s

impl

ified

(soi

l-stru

ctur

e in

tera

ctio

n an

alys

is n

ot re

quire

d) fo

r roc

k w

ith s

hear

wav

e ve

loci

ty >

2800

m/s

•Li

quef

actio

n is

not

an

issu

e fo

r NH

SB

; but

may

be

for a

t-gra

de s

truct

ures

on

soil

•R

egio

nal C

onsi

dera

tions

: S

oil o

ver s

hallo

w ro

ck s

ites

may

be

foun

d in

mos

t are

as o

f th

e U

S, i

nlan

d of

the

Atla

ntic

and

Gul

f Coa

stal

Pla

ins

and

away

from

the

high

er

upla

nd a

reas

.

Slid

e42

NG

NP

TE

AMA

ltern

ativ

e Si

te C

ondi

tions

(con

t.)

•D

eep

Soil

–Si

te is

cha

ract

eriz

ed b

y a

com

pete

nt s

oil b

earin

g la

yer w

ithin

60

m o

f gro

und

surfa

ce.

Roc

k is

muc

h de

eper

.–

Foun

datio

n m

at fo

r the

NH

SB w

ill be

on

the

com

pete

nt s

oil l

ayer

–O

ptim

al e

mbe

dmen

t dep

th w

ill te

nd to

be

dete

rmin

ed b

y th

e de

pth

to th

e co

mpe

tent

soi

l lay

er–

Soil

exca

vatio

n w

ill ha

ve m

ax s

ide

slop

es o

f 1.5

H:1

V or

2H

:1V

requ

iring

a m

uch

larg

er

exca

vatio

n as

the

foun

ding

laye

r bec

omes

dee

per (

or e

mbe

dmen

t is

grea

ter)

–M

inim

al o

r par

tial e

mbe

dmen

t may

be

pref

erre

d du

e to

:–

-low

er e

xcav

atio

n co

st–

-red

uced

nee

d fo

r gro

undw

ater

con

trol (

dew

ater

ing

and

wat

erpr

oofin

g)–

-red

uced

sto

rm w

ater

con

trol i

ssue

s du

ring

cons

truct

ion

–So

il liq

uefa

ctio

n du

ring

earth

quak

e m

ay b

e a

prob

lem

for N

HSB

and

at-g

rade

stru

ctur

es

–G

reat

er e

mbe

dmen

t dep

th m

ay p

rovi

de in

crea

sed

foun

datio

n be

arin

gca

paci

ty a

nd re

duce

d ris

k of

liqu

efac

tion

–Se

ism

ic d

esig

n w

ill re

quire

soi

l-stru

ctur

e in

tera

ctio

n an

alys

is–

Reg

iona

l Con

side

ratio

ns:

Dee

p so

il si

tes

are

typi

cal o

f the

Atla

ntic

and

Gul

f Coa

stal

Pla

ins,

an

d in

term

ount

ain

allu

vial

val

leys

(prim

arily

in th

e w

este

rn U

S)

Slid

e43

NG

NP

TE

AM

Rec

omm

enda

tions

•Th

e pa

rtial

em

bedm

ent s

chem

e sc

ores

bes

t for

bot

h R

ock

site

and

S

oil s

ite, a

nd w

ill th

eref

ore

be re

com

men

ded.

The

deg

ree

of p

artia

lem

bedm

ent w

ill be

opt

imiz

ed fo

r the

site

dur

ing

the

conc

eptu

al

desi

gn p

hase

. The

cas

e of

a s

oil s

ite w

ith a

dee

p w

ater

tabl

e is

not

show

n in

rank

ing

tabl

e bu

t it i

s ex

pect

ed th

e sc

ores

wou

ld b

e es

sent

ially

the

sam

e as

for a

rock

site

with

dee

p w

ater

tabl

e.•

The

PB

MR

des

ign

is q

uite

flex

ible

to a

ccom

mod

ate

vary

ing

site

ge

otec

hnic

al c

ondi

tions

. Acc

ess

to th

e re

acto

r bui

ldin

g ca

n be

ad

just

ed fo

r a g

iven

site

with

out a

sig

nific

ant i

mpa

ct o

n th

e la

yout

of

maj

or s

yste

ms

with

in th

e bu

ildin

g.

Slid

e44

NG

NP

TE

AM

Rol

e of

Rea

ctor

Em

bedm

ent i

n Sa

tisfy

ing

Rea

ctor

B

uild

ing

Func

tions

and

Req

uire

men

ts

•Th

e R

eact

or B

uild

ing

stud

y is

con

duct

ed in

two

parts

, Rea

ctor

B

uild

ing

Func

tions

and

Req

uire

men

ts a

nd R

eact

or E

mbe

dmen

t •

The

Rea

ctor

Bui

ldin

g Fu

nctio

ns a

nd R

equi

rem

ents

por

tion

of th

is

stud

y de

fines

reac

tor b

uild

ing

func

tions

and

requ

irem

ents

•

Func

tions

and

Req

uire

men

ts m

ust b

e fu

lfille

d in

depe

nden

tly o

f the

exte

nt o

f em

bedm

ent.

The

func

tions

and

requ

irem

ents

por

tion

of th

isst

udy

also

pro

vide

s as

a p

relim

inar

y se

t of l

icen

sing

bas

is e

vent

sin

volv

ing

leak

s an

d br

eaks

in th

e P

HTS

and

SH

TS p

ipin

g th

at th

e bu

ildin

g m

ust w

ithst

and.

The

func

tions

and

requ

irem

ents

sec

tion

also

eva

luat

es a

num

ber o

f alte

rnat

ive

desi

gn s

trate

gies

for

miti

gatin

g pi

pe b

reak

s an

d m

inim

izin

g ra

diol

ogic

al re

leas

es fr

omth

ebu

ildin

g. T

hese

opt

ions

can

all

be a

pplie

d to

any

leve

l of r

eact

orem

bedm

ent.

Hen

ce th

e fu

nctio

ns a

nd re

quire

men

ts s

ectio

n is

not

de

pend

ent o

n th

e ou

tcom

e or

con

clus

ions

of t

he re

acto

r em

bedm

ent

sect

ion.

Slid

e45

NG

NP

TE

AM

Rol

e of

Rea

ctor

Em

bedm

ent i

n Sa

tisfy

ing

Rea

ctor

B

uild

ing

Func

tions

and

Req

uire

men

ts (c

ont.)

•Th

e re

acto

r em

bedm

ent s

ectio

n is

som

ewha

t dep

ende

nt o

n th

e re

sults

of

func

tions

and

requ

irem

ents

sec

tion

but i

t is

conc

lude

d th

at c

onsi

dera

tions

that

hav

e be

en in

clud

ed in

the

reac

tor e

mbe

dmen

t sec

tion

are

to a

maj

or

exte

nt re

late

d to

ens

urin

g th

at fu

nctio

nal n

eeds

and

requ

irem

ents

iden

tifie

d in

func

tions

and

requ

irem

ents

sec

tion

can

be m

et. T

he re

acto

r em

bedm

ent

sect

ion

also

def

ines

a s

et o

f ext

erna

l haz

ards

and

phy

sica

l sec

urity

requ

irem

ents

that

the

build

ing

desi

gn m

ust f

ulfil

l, th

e ca

pabi

litie

s fo

r whi

ch

are

sign

ifica

ntly

influ

ence

d by

the

leve

l of e

mbe

dmen

t.•

A fu

rther

leve

l of s

yner

gy b

etw

een

func

tions

and

requ

irem

ents

sec

tion

and

the

reac

tor e

mbe

dmen

t sec

tion

is a

lso

prov

ided

by

the

use

of a

com

mon

set

of

crit

eria

for e

valu

atin

g th

e va

rious

des

ign

stra

tegi

es fo

r miti

gatin

g pi

pe

brea

ks in

the

func

tions

and

requ

irem

ents

sec

tion

and

the

eval

uatin

g th

e al

tern

ativ

e em

bedm

ent o

ptio

ns in

the

reac

tor e

mbe

dmen

t sec

tion.

The

seev

alua

tion

crite

ria a

re d

eriv

ed fr

om th

e fu

nctio

ns a

nd re

quire

men

ts s

ectio

n fu

nctio

nal r

equi

rem

ents

.

Slid

e46

NG

NP

TE

AM

Rol

e of

Rea

ctor

Em

bedm

ent i

n Sa

tisfy

ing

Rea

ctor

B

uild

ing

Func

tions

and

Req

uire

men

ts (c

ont.)

Func

tiona

l nee

ds to

be

met

by

the

reac

tor b

uild

ing

are

to:

•H

ouse

and

pro

vide

civ

il st

ruct

ural

sup

port

for t

he re

acto

r and

all

the

SSC

sw

ithin

the

scop

e of

the

NH

SS (

build

ing

geom

etry

/spa

ce)

–Th

is is

add

ress

ed b

y di

scus

sion

of f

ound

atio

n an

d se

ism

ic c

onsi

dera

tions

and

ope

ratio

n ne

eds

to h

ouse

S

SC

sw

ith p

rope

r orie

ntat

ion

with

resp

ect t

o th

e re

acto

r.•

Res

ist a

ll st

ruct

ural

load

s as

requ

ired

to s

uppo

rt bo

th re

quire

dan

d su

ppor

tive

safe

ty fu

nctio

ns

allo

cate

d to

the

NH

SB

–Th

is is

add

ress

ed b

y di

scus

sion

of f

ound

atio

n an

d se

ism

ic c

onsi

dera

tions

•Pr

otec

t the

SSC

sw

ithin

the

NH

SS th

at p

erfo

rm s

afet

y fu

nctio

ns fr

om a

ll in

tern

alan

d ex

tern

al

haza

rds

as id

entif

ied

in th

e LB

Esas

nec

essa

ry to

mee

t the

TLR

C–

This

is a

ddre

ssed

in d

efin

ition

of v

ario

us p

ipe

brea

k sc

enar

ios

in th

e fu

nctio

ns a

nd re

quire

men

ts s

ectio

n an

d th

e di

scus

sion

of

Nat

ural

and

Geo

logi

cal h

azar

ds in

The

reac

tore

mbe

dmen

t sec

tion

•Pr

ovid

e ph

ysic

al s

ecur

ity o

f vita

l are

as w

ithin

the

NH

SB a

gain

stac

ts o

f sab

otag

e an

d te

rroris

m–

This

is a

ddre

ssed

in th

e di

scus

sion

of M

alev

olen

t Haz

ards

of d

esig

n ba

sis

thre

ats

and

beyo

nd d

esig

n ba

sis

haza

rds

such

as

a no

n ac

cide

ntal

airc

raft

strik

e.•

Prov

ide

radi

onuc

lide

rete

ntio

n du

ring

blow

-dow

n an

d de

laye

d fu

el re

leas

e ph

ases

of

depr

essu

rizat

ion

even

ts.

–Th

is is

add

ress

ed in

the

disc

ussi

on o

f ope

ratio

nal r

equi

rem

ents

not

ing

the

leve

l of e

mbe

dmen

t can

affe

ct th

e ar

rang

emen

t of a

ven

t sta

ck if

requ

ired.

An

elev

ated

ven

t sta

ck is

not

acc

ount

ed fo

r in

dete

rmin

ing

dose

ra

tes

at th

e E

AB

. Gro

und

leve

l rel

ease

is a

ssum

ed fo

r all

alte

rnat

ive

sche

mes

.

Slid

e47

NG

NP

TE

AM

Bac

k U

p Sl

ides

•Sc

orin

g Ta

bles

–R

ock

Site

s (D

eep

Wat

er T

able

)•

Min

imal

Em

bedm

ent

•P

artia

l Em

bedm

ent

•Fu

ll E

mbe

dmen

t–

Soi

l Site

s (H

igh

Wat

er T

able

) •

Min

imal

Em

bedm

ent

•P

artia

l Em

bedm

ent

•Fu

ll E

mbe

dmen

t

Slid

e48

NG

NP

TE

AM

520

100

Tota

lWei

ght P

erce

nt

606

10C

onst

ruct

ion

Con

side

ratio

ns (N

ote

10)

606

10G

eote

chni

cal C

onst

rain

ts a

nd

Foun

datio

n P

erfo

rman

ce (N

ote

9)

505

10W

ater

Tab

le E

ffect

s (N

ote

8)

808

10C

ost B

enef

it C

onsi

dera

tion

(Not

e 7)

0C

ost a

nd C

onst

ruct

ion

Com

plex

ity

204

5C

hem

ical

Rel

ease

s an

d E

xplo

sion

s (N

ote

6)

404

10N

atur

al G

eolo

gica

l Phe

nom

ena

(Not

e 5)

604

15N

atur

al P

heno

men

a (N

ote

4)

604

15D

esig

n B

asis

Thr

eat a

nd M

alev

olen

t H

azar

ds (N

ote

3)

306

5H

eat D

issi

patio

n to

Env

ironm

ent (

Not

e 2)

0Sa

fety

, Inv

estm

ent P

rote

ctio

n, a

nd S

ecur

ity

606

10O

pera

tiona

l Nee

ds (S

afet

y eq

uipm

ent

layo

ut) (

Not

e 1)

0D

esig

n fo

r Ope

ratio

ns a

nd P

rope

r Sys

tem

Mec

hani

cal F

unct

ions

Roc

k M

inim

um E

mbe

dmen

t Alte

rnat

ive

Wei

ghte

d Sc

ore

Alte

rnat

ive

Satis

fact

ion

Scor

eW

eigh

tO

bjec

tive

Slid

e49

NG

NP

TE

AM

620

100

Tota

lWei

ght P

erce

nt

808

10C

onst

ruct

ion

Con

side

ratio

ns (N

ote

10)

505

10G

eote

chni

cal C

onst

rain

ts a

nd

Foun

datio

n P

erfo

rman

ce (N

ote

9)

505

10W

ater

Tab

le E

ffect

s (N

ote

8)

606

10C

ost B

enef

it C

onsi

dera

tion

(Not

e 7)

0C

ost a

nd C

onst

ruct

ion

Com

plex

ity

306

5C

hem

ical

Rel

ease

s an

d E

xplo

sion

s (N

ote

6)

606

10N

atur

al G

eolo

gica

l Phe

nom

ena

(Not

e 5)

906

15N

atur

al P

heno

men

a (N

ote

4)

906

15D

esig

n B

asis

Thr

eat a

nd M

alev

olen

t H

azar

ds (N

ote

3)

306

5H

eat D

issi

patio

n to

Env

ironm

ent (

Not

e 2)

0Sa

fety

, Inv

estm

ent P

rote

ctio

n, a

nd S

ecur

ity

808

10O

pera

tiona

l Nee

ds (S

afet

y eq

uipm

ent

layo

ut) (

Not

e 1)

0D

esig

n fo

r Ope

ratio

ns a

nd P

rope

r Sys

tem

Mec

hani

cal F

unct

ions

Roc

k Pa

rtia

lEm

bedm

ent A

ltern

ativ

e

Slid

e50

NG

NP

TE

AM

455

100

Tota

lWei

ght P

erce

nt

202

10C

onst

ruct

ion

Con

side

ratio

ns (N

ote

10)

404

10G

eote

chni

cal C

onst

rain

ts a

nd

Foun

datio

n P

erfo

rman

ce (N

ote

9)

505

10W

ater

Tab

le E

ffect

s (N

ote

8)

202

10C

ost B

enef

it C

onsi

dera

tion

(Not

e 7)

0C

ost a

nd C

onst

ruct

ion

Com

plex

ity

204

5C

hem

ical

Rel

ease

s an

d E

xplo

sion

s (N

ote

6)

303

10N

atur

al G

eolo

gica

l Phe

nom

ena

(Not

e 5)

906

15N

atur

al P

heno

men

a (N

ote

4)

105

715

Des

ign

Bas

is T

hrea

t and

Mal

evol

ent

Haz

ards

(Not

e 3)

306

5H

eat D

issi

patio

n to

Env

ironm

ent (

Not

e 2)

0Sa

fety

, Inv

estm

ent P

rote

ctio

n, a

nd S

ecur

ity

505

10O

pera

tiona

l Nee

ds (S

afet

y eq

uipm

ent

layo

ut) (

Not

e 1)

0D

esig

n fo

r Ope

ratio

ns a

nd P

rope

r Sys

tem

Mec

hani

cal F

unct

ions

Roc

k Fu

ll Em

bedm

ent A

ltern

ativ

e

Slid

e51

NG

NP

TE

AM

590

100

Tota

lWei

ght P

erce

nt

606

10C

onst

ruct

ion

Con

side

ratio

ns (N

ote

10)

808

10G

eote

chni

cal C

onst

rain

ts a

nd

Foun

datio

n P

erfo

rman

ce (N

ote

9)

100

1010

Wat

er T

able

Effe

cts

(Not

e 8)

808

10C

ost B

enef

it C

onsi

dera

tion

(Not

e 7)

0C

ost a

nd C

onst

ruct

ion

Com

plex

ity

204

5C

hem

ical

Rel

ease

s an

d E

xplo

sion

s (N

ote

6)

404

10N

atur

al G

eolo

gica

l Phe

nom

ena

(Not

e 5)

604

15N

atur

al P

heno

men

a (N

ote

4)

604

15D

esig

n B

asis

Thr

eat a

nd M

alev

olen

t H

azar

ds (N

ote

3)

306

5H

eat D

issi

patio

n to

Env

ironm

ent (

Not

e 2)

0Sa

fety

, Inv

estm

ent P

rote

ctio

n, a

nd S

ecur

ity

606

10O

pera

tiona

l Nee

ds (S

afet

y eq

uipm

ent

layo

ut) (

Not

e 1)

0D

esig

n fo

r Ope

ratio

ns a

nd P

rope

r Sys

tem

Mec

hani

cal F

unct

ions

Soil

Min

imum

Em

bedm

ent A

ltern

ativ

e

Wei

ghte

d Sc

ore

Alte

rnat

ive

Satis

fact

ion

Scor

eW

eigh

tO

bjec

tive

Slid

e52

NG

NP

TE

AM

650

100

Tota

lWei

ght P

erce

nt

808

10C

onst

ruct

ion

Con

side

ratio

ns (N

ote

10)

606

10G

eote

chni

cal C

onst

rain

ts a

nd

Foun

datio

n P

erfo

rman

ce (N

ote

9)

707

10W

ater

Tab

le E

ffect

s (N

ote

8)

606

10C

ost B

enef

it C

onsi

dera

tion

(Not

e 7)

0C

ost a

nd C

onst

ruct

ion

Com

plex

ity

306

5C

hem

ical

Rel

ease

s an

d E

xplo

sion

s (N

ote

6)

606

10N

atur

al G

eolo

gica

l Phe

nom

ena

(Not

e 5)

906

15N

atur

al P

heno

men

a (N

ote

4)

906

15D

esig

n B

asis

Thr

eat a

nd M

alev

olen

t H

azar

ds (N

ote

3)

306

5H

eat D

issi

patio

n to

Env

ironm

ent (

Not

e 2)

0Sa

fety

, Inv

estm

ent P

rote

ctio

n, a

nd S

ecur

ity

808

10O

pera

tiona

l Nee

ds (S

afet

y eq

uipm

ent

layo

ut) (

Not

e 1)

0D

esig

n fo

r Ope

ratio

ns a

nd P

rope

r Sys

tem

Mec

hani

cal F

unct

ions

Soil

Par

tialE

mbe

dmen

t Alte

rnat

ive

Slid

e53

NG

NP

TE

AM

385

100

Tota

lWei

ght P

erce

nt

202

10C

onst

ruct

ion

Con

side

ratio

ns (N

ote

10)

202

10G

eote

chni

cal C

onst

rain

ts a

nd

Foun

datio

n P

erfo

rman

ce (N

ote

9)

00

10W

ater

Tab

le E

ffect

s (N

ote

8)

202

10C

ost B

enef

it C

onsi

dera

tion

(Not

e 7)

0C

ost a

nd C

onst

ruct

ion

Com

plex

ity

204

5C

hem

ical

Rel

ease

s an

d E

xplo

sion

s (N

ote

6)

303

10N

atur

al G

eolo

gica

l Phe

nom

ena

(Not

e 5)

906

15N

atur

al P

heno

men

a (N

ote

4)

105

715

Des

ign

Bas

is T

hrea

t and

Mal

evol

ent

Haz

ards

(Not

e 3)

306

5H

eat D

issi

patio

n to

Env

ironm

ent (

Not

e 2)

0Sa

fety

, Inv

estm

ent P

rote

ctio

n, a

nd S

ecur

ity

505

10O

pera

tiona

l Nee

ds (S

afet

y eq

uipm

ent

layo

ut) (

Not

e 1)

0D

esig

n fo

r Ope

ratio

ns a

nd P

rope

r Sys

tem

Mec

hani

cal F

unct

ions

Soil

Full

Embe

dmen

t Alte

rnat

ive

Slid

e2

PB

MR

TE

AM

Dis

cuss

ion

Topi

cs•

Intr

oduc

tion

–O

bjec

tives

and

Sco

pe–

App

roac

h–

Rel

atio

nshi

p w

ith R

eact

or E

mbe

dmen

t •

Rev

iew

of F

unct

ions

and

Req

uire

men

ts•

Eval

uatio

n of

Alte

rnat

ive

Bui

ldin

g D

esig

n St

rate

gies

–Ev

alua

tion

Crit

eria

–D

efin

ition

of A

ltern

ativ

e D

esig

n St

rate

gies

–Ev

alua

tion

of R

adio

nucl

ide

Ret

entio

n an

d Pr

essu

re C

apac

ity–

Eval

uatio

n of

Cos

ts–

Inte

grat

ed E

valu

atio

n vs

. Fun

ctio

nal R

equi

rem

ents

–A

lloca

tion

of R

adio

nucl

ide

Ret

entio

n C

apab

ility

•O

pen

Issu

es a

nd A

dditi

onal

R&

D a

nd E

ngin

eerin

g St

udie

s•

Con

clus

ions

and

Rec

omm

enda

tions

Slid

e3

PB

MR

TE

AM

Ove

rvie

w

•S

tudy

stru

ctur

ed in

to tw

o co

mpl

emen

tary

sec

tions

:–

Rea

ctor

Bui

ldin

g Te

chni

cal a

nd F

unct

iona

l Req

uire

men

ts•

Ove

rall

requ

irem

ents

for a

ll de

sign

alte

rnat

ives

•D

efin

ition

of p

relim

inar

y lic

ensi

ng b

asis

eve

nts

with

focu

s on

HP

Ble

aks

and

brea

ks•

Eva

luat

ion

of a

ltern

ativ

e de

sign

app

roac

hes

to m

eetin

g re

acto

r bu

ildin

g sa

fety

func

tions

inde

pend

ent o

f em

bedm

ent

–E

valu

atio

n of

Rea

ctor

Bui

ldin

g E

mbe

dmen

t•

Def

initi

on o

f pre

limin

ary

licen

sing

bas

is e

vent

s in

volv

ing

exte

rnal

haza

rds

and

phys

ical

sec

urity

thre

ats

•E

valu

atio

n of

alte

rnat

ive

appr

oach

es to

reac

tor e

mbe

dmen

t in

depe

nden

t of r

eact

or b

uild

ing

desi

gn a

ltern

ativ

es

•In

tegr

ated

eva

luat

ion

and

reco

mm

enda

tions

for

cons

ider

atio

n in

Con

cept

ual D

esig

n of

NG

NP

RB

Slid

e4

PB

MR

TE

AM

Key

Con

trib

utor

s to

Rea

ctor

Bui

ldin

g Se

ctio

n

•P

BM

R:

–R

oger

You

ng, D

irk U

ys, N

ilen

Nai

du, J

ohan

Stra

uss,

Wilm

a va

n E

ck–

Sha

w:

–P

eter

Wel

ls, B

ob W

ilmer

, Nic

hola

s M

enou

nos,

Mic

hael

D

avid

son

•Te

chno

logy

Insi

ghts

: –

Kar

l Fle

min

g, F

red

Sila

dy, D

ave

Dilli

ng, F

red

Torr

i, Lo

ri M

asca

ro

•W

estin

ghou

se:

–Ji

m W

inte

rs

Slid

e5

PB

MR

TE

AM

Obj

ectiv

es a

nd S

cope

of

Rea

ctor

Bui

ldin

g Se

ctio

n

•O

bjec

tives

:–

To d

evel

op T

echn

ical

and

Fun

ctio

nal R

equi

rem

ents

(T&

FRs)

for t

he N

GN

P re

acto

r bui

ldin

g,

–To

iden

tify

and

eval

uate

alte

rnat

ive

reac

tor b

uild

ing

desi

gn

stra

tegi

es

•S

cope

:–

Def

initi

on o

f ove

rall

tech

nica

l req

uire

men

ts fo

r the

reac

tor

build

ing

–E

valu

atio

n of

des

ign

optio

ns to

mee

ting

thes

e re

quire

men

ts–

Allo

catio

n of

radi

olog

ical

rete

ntio

n re

quire

men

ts fo

r rea

ctor

bu

ildin

g

Slid

e6

PB

MR

TE

AM

RB

Req

uire

men

tsA

ppro

ach

Slid

e7

PB

MR

TE

AM

Scop

e of

Rea

ctor

Bui

ldin

g

Rea

ctor

Bui

ldin

g

Slid

e8

PB

MR

TE

AMM

ajor

Con

tent

s of

Rea

ctor

Bui

ldin

g

•R

eact

or U

nit S

yste

m,

•C

ore

Con

ditio

ning

Sys

tem

, •

Rea

ctor

Cav

ity C

oolin

g S

yste

m,

•Fu

el H

andl

ing

and

Sto

rage

Sys

tem

, •

Hel

ium

Ser

vice

Sys

tem

, •

Hea

ting

Ven

tilat

ion

and

Air

Con

ditio

ning

(HV

AC

)•

Pre

ssur

e R

elie

f Sys

tem

•

Oth

er N

HS

S s

uppo

rt sy

stem

s.•

Por

tion

of th

e H

eat T

rans

port

Sys

tem

insi

de th

e bu

ildin

g–

Prim

ary

Hea

t Tra

nspo

rt S

yste

m–

IHX

ves

sels

, circ

ulat

ors

and

–Pa

rt of

Sec

onda

ry H

eat T

rans

port

Syst

em

Slid

e9

PB

MR

TE

AM

NG

NP

PBM

R S

afet

y D

esig

n A

ppro

ach

•P

BM

R s

afet

y ph

iloso

phy

star

ts w

ith th

e ap

plic

atio

n of

de

fens

e-in

-dep

th p

rinci

ples

in m

akin

g fu

ndam

enta

l de

sign

sel

ectio

ns:

–In

here

nt re

acto

r cha

ract

eris

tics

–M

ultip

le, r

obus

t, an

d co

ncen

tric

barri

ers

to ra

dion

uclid

e re

leas

e.–

Con

serv

ativ

e de

sign

app

roac

hes

to s

uppo

rt st

able

pla

nt

oper

atio

n w

ith la

rge

mar

gins

to s

afet

y lim

its

–In

the

even

t saf

ety

func

tions

are

cha

lleng

ed, t

he s

afet

y ph

iloso

phy

is to

util

ize

the

inhe

rent

reac

tor c

hara

cter

istic

s an

dpa

ssiv

e de

sign

feat

ures

to fu

lfill

the

requ

ired

safe

ty fu

nctio

ns.

–A

dditi

onal

act

ive

engi

neer

ed s

yste

ms

and

oper

ator

act

ions

are

pr

ovid

ed to

redu

ce th

e ch

alle

nge

to p

lant

saf

ety

and

prov

ide

defe

nse-

in-d

epth

in p

reve

ntin

g an

d m

itiga

ting

acci

dent

s.–

Avo

id n

eed

for e

arly

ope

rato

r int

erve

ntio

n, o

r ear

ly fu

nctio

ning

ofan

y ac

tive

syst

ems

to m

aint

ain

safe

sta

ble

stat

e.

Slid

e10

PB

MR

TE

AM

PBM

R N

GN

P A

ppro

ach

toPl

ant C

apab

ility

Def

ense

-in-D

epth

Fuel

Bar

rier

Coo

lant

Pre

ssur

e B

ound

ary

Rea

ctor

Bui

ldin

g B

arrie

r

Cor

e In

vent

ory

Of R

adio

activ

eM

ater

ial

Pass

ive

Engi

neer

edFe

atur

es a

nd S

SCs

Act

ive

Engi

neer

edFe

atur

es a

nd S

SCs

Inhe

rent

Pro

pert

ies

of C

ore,

Fue

l, M

oder

ator

, C

oola

nt, a

nd R

eact

or V

esse

l

Safe

ty F

unct

ions

Prot

ectin

g B

arrie

rs

Site

Sel

ectio

n

Fuel

Bar

rier

Coo

lant

Pre

ssur

e B

ound

ary

Rea

ctor

Bui

ldin

g B

arrie

r

Cor

e In

vent

ory

Of R

adio

activ

eM

ater

ial

Pass

ive

Engi

neer

edFe

atur

es a

nd S

SCs

Act

ive

Engi

neer

edFe

atur

es a

nd S

SCs

Inhe

rent

Pro

pert

ies

of C

ore,

Fue

l, M

oder

ator

, C

oola

nt, a

nd R

eact

or V

esse

l

Safe

ty F

unct

ions

Prot

ectin

g B

arrie

rs

Site

Sel

ectio

n

Slid

e11

PB

MR

TE

AM

Rea

ctor

Bui

ldin

g Fu

nctio

ns

•S

afet

y Fu

nctio

ns–

Req

uire

d Sa

fety

Fun

ctio

nsar

e th

ose

func

tions

that

are

nec

essa

ry

and

suffi

cien

t to

mee

t the

dos

e lim

its fo

r Des

ign

Basi

s Ev

ents

and

dete

rmin

istic

ally

sel

ecte

d D

esig

n B

asis

Acc

iden

ts–

Supp

ortiv

e Sa

fety

Fun

ctio

nsar

e al

l oth

er fu

nctio

ns th

at c

ontri

bute

to

the

prev

entio

n or

miti

gatio

n of

acc

iden

ts a

nd s

uppo

rt th

e pl

ant

capa

bilit

ies

for d

efen

se-in

-dep

th•

Phy

sica

l Sec

urity

Fun

ctio

ns–

Pro

tect

the

reac

tor a

nd v

ital e

quip

men

t fro

m d

esig

n ba

sis

thre

ats

asso

ciat

ed w

ith a

cts

of s

abot

age

and

terro

rism

•O

ther

Fun

ctio

ns–

Func

tions

nec

essa

ry fo

r pla

nt c

onst

ruct

ion,

ope

ratio

n, m

aint

enan

ce,

acce

ss, i

nspe

ctio

n, w

orke

r pro

tect

ion,

and

con

trol o

f rou

tine

rele

ases

of ra

dioa

ctiv

e m

ater

ial d

urin

g no

rmal

ope

ratio

n

Slid

e12

PB

MR

TE

AM

RB

Fun

ctio

nal R

equi

rem

ents

•Th

e sc

ope

of fu

nctio

ns a

nd re

quire

men

ts fo

r the

reac

tor b

uild

ing

are

allo

cate

d in

the

PC

DR

to :

–N

ucle

ar H

eat S

uppl

y B

uild

ing

(NH

SB

) –

NH

SS H

VAC

Sys

tem

–N

HSB

Pre

ssur

e R

elie

f Sys

tem

(PR

S)•

The

RB

sha

ll pe

rform

the

follo

win

g re

quire

d fu

nctio

ns:

–H

ouse

and

pro

vide

civ

il st

ruct

ural

sup

port

for t

he re

acto

r and

all

the

SSC

s w

ithin

the

scop

e of

the

NH

SS (

build

ing

geom

etry

/spa

ce)

–R

esis

t all

stru

ctur

al lo

ads

as re

quire

d fo

r req

uire

d sa

fety

func

tions

allo

cate

d to

the

NH

SS

–P

rote

ct th

e S

SC

s w

ithin

the

NH

SS

that

per

form

requ

ired

safe

ty

func

tions

from

all

inte

rnal

and

ext

erna

l haz

ards

as

iden

tifie

d in

the

LBE

s–

Prov

ide

phys

ical

sec

urity

of v

ital a

reas

of t

he N

HSS

aga

inst

act

s of

sa

bota

ge a

nd te

rroris

m•

The

RB

sha

ll pe

rform

sup

porti

ve s

afet

y fu

nctio

ns in

volv

ing

the

rete

ntio

n of

radi

onul

cide

s re

leas

ed fr

om th

e P

HTS

HP

B a

nd th

e lim

iting

of a

ir in

gres

s fo

llow

ing

HP

B le

aks

and

brea

ks

Slid

e13

PB

MR

TE

AM

PRS

Func

tiona

l Req

uire

men

ts

•Th

e P

ress

ure

Rel

ief S

yste

m s

hall

perfo

rm th

e fo

llow

ing

func

tions

:–

The

PR

S m

ust b

e co

mpa

tible

with

the

NH

SB

bou

ndar

y fu

nctio

nal r

equi

rem

ent.

–Th

e P

RS

sha

ll be

des

igne

d, a

nd th

e N

HS

B c

ompa

rtmen

ts

shal

l be

size

d so

that

leak

s an

d br

eaks

up

to 1

0mm

eq

uiva

lent

bre

ak s

ize

anyw

here

alo

ng th

e P

HTS

HP

B d

oes

not o

pen

the

PR

S s

o th

at H

VA

C fi

ltrat

ion

shal

l be

mai

ntai

ned

–G

ener

ally

, PR

S m

ust o

pen

to p

reve

nt o

verp

ress

ure

and

dam

age

to s

afet

y re

late

d S

SC

s, a

nd re

clos

e to

ena

ble

post

-bl

ow-d

own

filtra

tion

and

leak

age

cont

rol f

or th

e fo

llow

ing

desi

gn b

asis

eve

nt c

ondi

tions

.•

Brea

ks in

the

PHTS

HPB

in e

ach

NH

SB c

ompa

rtmen

t con

tain

ing

NH

SS p

ipin

g or

ves

sels

up

to 1

00m

m e

quiv

alen

t bre

ak s

ize

•Br

eaks

in th

e SH

TS H

PB in

eac

h N

HSB

com

partm

ent c

onta

inin

g N

HSS

pip

ing

or v

esse

ls u

p to

1m

eter

equ

ival

ent b

reak

siz

e

Slid

e14

PB

MR

TE

AM

HVA

C F

unct

iona

l Req

uire

men

ts

•Th

e N

HS

S H

VA

C S

yste

m s

hall

perfo

rm th

e fo

llow

ing

func

tions

:–

The

HV

AC

sys

tem

sha

ll•

mai

ntai

n in

tern

al e

nviro

nmen

tal c

ondi

tions

for N

HSS

SSC

s du

ring

norm

al o

pera

tion

•M

aint

ain

nega

tive

pres

sure

in th

e R

BV

V a

reas

to k

eep

norm

al

rele

ases

ALA

RA

•su

rviv

e co

nditi

ons

for a

ll A

OO

san

d D

BE

s, a

nd

•pr

ovid

e a

post

-eve

nt c

lean

up fu

nctio

n fo

r all

AO

Os

and

DB

Es.

–Th

e H

VA

C s

yste

m m

ayal

so p

lay

a ro

le in

miti

gatin

g th

e re

leas

e of

radi

oact

ive

mat

eria

l dep

endi

ng o

n th

e se

lect

ed

desi

gn s

trate

gy

Slid

e15

PB

MR

TE

AM

Phys

ical

Sec

urity

Req

uire

men

ts

•In

sigh

ts fr

om e

xper

ienc

e w

ith A

P10

00 b

eing

use

d to

est

ablis

h ph

ysic

al s

ecur

ity re

quire

men

ts fo

r the

RB

•S

ecur

ity re

quire

men

ts e

stab

lishe

d in

10

CFR

50,

52,

and

73

•N

eed

to d

efin

e “v

ital a

reas

”and

“vita

l equ

ipm

ent”

in li

ght o

f NG

NP

ris

k-in

form

ed p

erfo

rman

ce-b

ased

lice

nsin

g ap

proa

ch a

nd S

SC

sa

fety

cla

ssifi

catio

n m

etho

d•

Vita

l are

as s

ubje

cted

to ri

goro

us a

naly

sis

to a

ddre

ss c

lass

ified

Des

ign

Bas

is T

hrea

ts w

hich

invo

lve

both

des

ign

basi

s an

d be

yond

de

sign

bas

is a

ccid

ent c

ondi

tions

•R

obus

t phy

sica

l pro

tect

ion

prov

ided

in d

esig

n to

add

ress

DB

T ex

pect

ed to

con

tribu

te to

mar

gins

in th

e ca

paci

ty to

with

stan

d lo

ads

from

inte

rnal

and

ext

erna

l eve

nt a

ccid

ents

•C

onsi

dera

tion

of th

ese

requ

irem

ents

not

exp

ecte

d to

influ

ence

ev

alua

tion

of re

acto

r bui

ldin

g co

nfin

emen

t opt

ions

but

cou

ld

influ

ence

eva

luat

ion

of e

mbe

dmen

t

Slid

e16

PB

MR

TE

AMLi

cens

ing

Bas

is E

vent

sFr

amin

g R

eact

or B

uild

ing

Req

uire

men

ts

Slid

e17

PB

MR

TE

AM

NG

NP

Lice

nsin

g B

asis

Eve

nts

•LB

Es

to b

e de

fined

via

full

scop

e P

RA

dur

ing

conc

eptu

al d

esig

n,

•Fo

r thi

s st

udy

LBE

sar

e de

fined

by

engi

neer

ing

judg

men

t bas

ed o

n re

view

s of

the

plan

t des

ign

per P

CD

R a

nd e

xper

ienc

e w

ith H

TGR

and

spec

ifica

lly P

BM

R P

RA

s•

Cat

egor

ies

of L

BE

sin

clud

e–

Inte

rnal

Eve

nts

•H

PB

leak

s an

d br

eaks

in P

HTS

and

SH

TS•

Tran

sien

ts w

ith a

nd w

ithou

t rea

ctiv

ity a

dditi

on–

Inte

rnal

Pla

nt H

azar

d Ev

ents

•In

tern

al fi

res

and

flood

s•

Rot

atin

g m

achi

nery

mis

sile

s•

Hig

h en

ergy

line

and

tank

bre

aks

–Ex

tern

al P

lant

Haz

ard

Even

ts

•H

ydro

gen

Pro

cess

eve

nts

•S

eism

ic E

vent

s•

Airc

raft

cras

hes

and

trans

porta

tion

acci

dent

s•

Hig

h w

inds

and

win

d ge

nera

ted

mis

sile

s

Slid

e18

PB

MR

TE

AM

Ass

umed

LBEs

Invo

lvin

g Le

aks

and

Bre

aks

•A

OO

sas

sum

ed to

incl

ude

leak

s up

to 1

0mm

in b

reak

siz

e an

ywhe

re a

long

the

PH

TS a

nd S

HTS

HP

B in

side

the

RB

–P

RS

and

HV

AC

des

igne

d to

kee

p H

VA

C ru

nnin

g du

ring

such

leak

s•

DB

Es

are

expe

cted

incl

ude:

–U

p to

50m

m b

reak

s in

the

FHS

S p

ipes

abo

ve o

r bel

ow th

e R

PV

–U

p to

100

mm

bre

aks

in th

e C

IP, C

OP

or o

ther

PH

TS p

ipin

g ab

ove

orbe

low

the

core

incl

udin

g pi

pe to

ves

sel n

ozzl

e w

elds

–B

reak

s in

HS

S p

ipin

g bo

unde

d by

100

m b

reak

siz

e–

Up

to 1

met

er (S

EG

B) b

reak

s in

the

SH

TS p

ipin

g in

side

the

NH

SB

–R

equi

rem

ent i

s to

mai

ntai

n st

ruct

ural

inte

grity

and

avo

id d

amag

eto

any

saf

ety

rela

ted

SS

C in

side

the

NH

SB

•B

DB

Esar

e ex

pect

ed to

incl

ude

–Le

aks

and

smal

l bre

aks

in th

e R

PV

–S

imul

tane

ous

smal

l lea

ks a

bove

and

bel

ow th

e co

re–

Up

to 1

met

er (S

EG

B) b

reak

s in

the

PH

TS–

Req

uire

men

t is

to m

eet T

LRC

usi

ng re

alis

tic a

ssum

ptio

ns–

Stru

ctur

al in

tegr

ity o

f the

RP

V a

nd m

ajor

PH

TS S

SC

s m

aint

aine

d

Slid

e19

PB

MR

TE

AM

HPB

leak

s an

d B

reak

s Li

cens

ing

Bas

is E

vent

s

> 10

0 to

1,0

00*

Larg

e<1

0-4

Bey

ond

Des

ign

Bas

is E

vent

s

> 10

to 1

00M

ediu

m10

-4to

10-

2D

esig

n B

asis

E

vent

s

1 to

10

Sm

all

=10

-2A

ntic

ipat

edO

pera

tiona

lO

ccur

renc

e

Equ

ival

ent B

reak

S

ize

(mm

) B

reak

Cat

egor

yFr

eque

ncy

Ran

ge

(per

pla

nt y

ear

LBE

Type

*cor

resp

onds

to a

dou

ble

ende

d gu

illotin

e br

eak

of a

700

mm

pip

e

Slid

e20

PB

MR

TE

AM

NH

SB R

equi

rem

ents

Sum

mar

y

•P

lant

Per

form

ance

Req

uire

men

ts–

Plan

t rel

iabi

lity

and

capa

city

fact

or–

Plan

t inv

estm

ent r

isk

•S

afet

y an

d Li

cens

ing

Req

uire

men

ts–

Top

Leve

l Reg

ulat

ory

Crit

eria

–Ap

plic

able

exi

stin

g N

RC

gui

danc

e–

Polic

ies

and

guid

ance

spe

cific

to a

dvan

ced

reac

tors

–Sa

fety

and

lice

nsin

g re

quire

men

ts k

eyed

to s

peci

fic c

ateg

orie

s of

LBEs

•Le

aks

and

brea

ks in

HP

B•

Tran

sien

ts•

Inte

rnal

fire

s an

d flo

ods

•S

eism

ic e

vent

s•

Bey

ond

desi

gn b

asis

airc

raft

cras

hes

and

larg

e fir

es a

nd e

xplo

sion

s•

Hyd

roge

n pr

oces

s ha

zard

s•

Phy

sica

l Sec

urity

Req

uire

men

ts

•P

lant

Cos

ts–

Cap

ital c

osts

–O

pera

ting

and

mai

nten

ance

cos

ts–

Tech

nolo

gy d

evel

opm

ent c

osts

•N

GN

P Sc

hedu

le•

Pub

lic A

ccep

tanc

e

Slid

e21

PB

MR

TE

AM

EVA

LUA

TIO

N O

F A

LTER

NA

TIVE

R

EAC

TOR

BU

ILD

ING

DES

IGN

STR

ATE

GIE

S(O

ther

than

Em

bedm

ent)

Slid

e22

PB

MR

TE

AM

App

roac

h to

Eva

luat

ion

of

Alte

rnat

ive

RB

Con

figur

atio

ns

•Id

entif

y re

quire

d an

d su

ppor

tive

safe

ty fu

nctio

ns o

f RB

•Id

entif

y al

tern

ativ

e de

sign

stra

tegi

es to

mee

t eac

h sa

fety

func

tion

•D

efin

e di

scre

te s

et o

f alte

rnat

ive

RB

des

ign

conf

igur

atio

ns w

ithap

prop

riate

com

bina

tions

of d

esig

n fe

atur

es•

Eva

luat

e ea

ch c

onfig

urat

ion

agai

nst t

echn

ical

and

func

tiona

l pe

rform

ance

crit

eria

–N

orm

al O

pera

ting

Func

tiona

l Req

uire

men

ts–

Inve

stm

ent P

rote

ctio

n Fu

nctio

nal R

equi

rem

ents

–S

afet

y Fu

nctio

nal R

equi

rem

ents

–P

hysi

cal S

ecur

ity /

Airc

raft

Cra

sh R

equi

rem

ents

–C

apita

l and

O&

M C

ost I

mpa

ct–

Lice

nsab

ility

•Id

entif

y ba

sis

for r

ecom

men

ding

one

or m

ore

alte

rnat

ive

conf

igur

atio

ns•

Res

ult i

s a

reco

mm

enda

tion

for t

he C

once

ptua

l Des

ign

to c

onsi

deri

nm

eetin

g th

e fu

nctio

nal a

nd te

chni

cal r

equi

rem

ents

and

not

a p

ropo

sed

desi

gn d

ecis

ion

Slid

e23

PB

MR

TE

AM

RB

Stu

dy E

valu

atio

n C

riter

ia a

nd W

eigh

tsC

riter

iaR

elat

ive

Wei

ght

•N

orm

al O

pera

tion

Req

uire

men

ts

10%

•In

vest

men

t Pro

tect

ion

Req

uire

men

t5%

•Sa

fety

Fun

ctio

nal R

equi

rem

ents

35

%–

HPB

leak

s/br

eaks

20%

–Se

ism

ic5%

–H

ydro

gen/

proc

ess

haza

rds

10%

•Se

curit

y /a

ircra

ft cr

ash

10%

•C

apita

l and

Ope

ratin

g C

ost

25%

•Li

cens

abili

ty15

%10

0%

Slid

e24

PB

MR

TE

AM

Ran

ge o

f Opt

ions

to M

eet S

truc

tura

l Bui

ldin

g Fu

nctio

ns

•A

ltern

ativ

e pr

essu

re re

lief a

ppro

ache

s–

Ope

n ve

nt p

athw

ays

–B

low

out

pan

els

–B

uild

ing

shap

e•

Pro

tect

ion

agai

nst i

nter

nal h

azar

ds–

Zoni

ng s

trate

gies

–In

tern

al b

arrie

rs•

Pro

tect

ion

agai

nst e

xter

nal h

azar

ds a

nd p

hysi

cal

secu

rity

thre

ats

–R

ound

vs.

rect

angu

lar s

hape

–R

eact

or e

mbe

dmen

t

Slid

e25

PB

MR

TE

AM

Ran

ge o

f Opt

ions

to

Miti

gate

Rel

ease

s an

d A

ir In

gres

s

•R

adio

nucl

ide

rete

ntio

n–

HV

AC

pre

ssur

e zo

nes

–H

VA

C fi

ltrat

ion

•A

ctiv

e•

Pass

ive

–C

ontro

lled

rele

ase

path

way

–B

uild

ing

leak

rate

con

trol

•A

ir in

gres

s co

ntro

l–

Bui

ldin

g le

ak ra

te c

ontro

l–

Re-

clos

able

ven

t dam

pers

–In

ert g

as in

ject

ion

not c

onsi

dere

d be

caus

e ca

pabi

lity

to in

ject

H

e or

N2

into

reac

tor c

ore

is a

lread

y in

here

nt in

refe

renc

e P

BM

R d

esig

n

Slid

e26

PB

MR

TE

AMRan

ge o

f Opt

iona

l Des

ign

App

roac

hes

for

Rad

ionu

clid

e R

eten

tion

and

Air

Ingr

ess

Miti

gatio

n

•U

nfilt

ered

Ven

ted

•Fi

ltere

d Ve

nted

w/ C

ontr

olle

d le

akag

e–

Opt

ion

with

del

ayed

rele

ase

filte

r cap

abili

ty–

Opt

ion

with

cap

abili

ty to

filte

r im

med

iate

and

de

laye

d ph

ases

of r

elea

se–

Opt

ions

with

diff

eren

t ven

t are

a vo

lum

es•

Pres

sure

Ret

aini

ng–

Opt

ions

with

diff

eren

t ven

t are

a vo

lum

es

Slid

e27

PB

MR

TE

AM

Alte

rnat

ive

App

roac

hes

to

Mee

ting

Req

uire

d R

B S

afet

y Fu

nctio

ns

•R

obus

t Stru

ctur

al d

esig

n•

Phy

sica

l sep

arat

ion

of S

SC

sP

reve

nt d

amag

e to

fuel

sto

rage

and

sa

fety

rela

ted

SS

Cs

from

HP

B b

reak

s

•R

obus

t stru

ctur

al d

esig

n•

Em

bedm

ent

Pro

vide

stru

ctur

al in

tegr

ity in

resp

onse

to

sei

smic

and

ext

erna

l eve

nt lo

ads

and

mal

evol

ent t

hrea

ts

•R

obus

t stru

ctur

al d

esig

n •

Larg

e op

en v

ent p

athw

ays

•R

uptu

re p

anel

s•

Add

ed e

xpan

sion

vol

ume

Pro

vide

stru

ctur

al in

tegr

ity in

resp

onse

to

HP

B b

reak

pre

ssur

e an

d te

mpe

ratu

re

load

s

Alte

rnat

ive

Des

ign

App

roac

hes

RB

Req

uire

d Sa

fety

Fun

ctio

n

Slid

e28

PB

MR

TE

AM

Alte

rnat

ive

App

roac

hes

to

Mee

ting

Supp

ortiv

e R

B S

afet

y Fu

nctio

ns

•Is

olat

ion

dam

pers

•B

last

pan

els

•P

hysi

cal s

epar

atio

n of

HV

AC

SS

Cs

Pre

vent

dam

age

to H

VA

C d

urin

g bl

ow-

dow

n fro

m H

PB

bre

aks

•H

VA

C w

ith a

ctiv

e fil

tratio

n •

pres

sure

and

radi

atio

n zo

nes

Mai

ntai

n en

viro

nmen

tal c

ondi

tions

and

co

ntro

l rad

ionu

clid

e re

leas

es d

urin

g no

rmal

pla

nt o

pera

tion

•Li

mit

RB

leak

rate

•P

rovi

sion

s fo

r ine

rting

of P

HTS

•P

rovi

sion

s fo

r ine

rting

of R

B

Lim

it po

tent

ial f

or a

ir in

gres

s to

PH

TS

follo

win

g la

rge

or m

ultip

le H

PB

bre

aks

•P

assi

ve fi

ltrat

ion

•R

e-es

tabl

ish

HV

AC

filtr

atio

n•

Con

trol r

elea

se p

athw

ay•

Lim

it R

B le

ak ra

te

Pro

vide

radi

onuc

lide

rete

ntio

n du

ring

dela

yed

fuel

rele

ase

phas

e

•P

assi

ve fi

ltrat

ion

•Li

mit

RB

leak

rate

Pro

vide

radi

onuc

lide

rete

ntio

n du

ring

blow

-dow

n ph

ase

Alte

rnat

ive

Des

ign

App

roac

hes

RB

Sup

porti

ve S

afet

y Fu

nctio

n

Slid

e29

PB

MR

TE

AM

Des

ign

Feat

ures

Com

mon

to a

ll Se

lect

ed

Alte

rnat

ive

Con

figur

atio

ns•

Rob

ust s

truct

ural

des

ign

to p

rote

ct a

gain

st lo

ads

from

inte

rnal

and

ext

erna

l ha

zard

s in

clud

ing

seis

mic

eve

nts,

hyd

roge

n pr

oces

s ha

zard

s an

d ph

ysic

alse

curit

y ag

ains

t mal

evol

ent a

cts

as d

escr

ibed

in s

elec

ted

licen

sing

bas

is

even

ts•

HVA

C fi

ltrat

ion

syst

em p

rovi

ded

to m

aint

ain

envi

ronm

enta

l con

ditio

ns a

nd

prov

ide

radi

onuc

lide

rete

ntio

n AL

ARA

durin

g no

rmal

ope

ratio

n an

dfo

ran

ticip

ated

ope

ratio

nal o

ccur

renc

es (A

OO

s)•

Bla

st p

anel

s or

isol

atio

n de

vice

s pr

ovid

ed to

isol

ate/

pro

tect

nor

mal

oper

atio

n H

VA

C fr

om H

PB

bre

aks

•P

hysi

cal i

sola

tion/

sepa

ratio

n pr

ovid

ed fo

r fue

l sto

rage

and

saf

ety

rela

ted

SS

Cs

from

blo

w-d

own

vent

pat

h fo

llow

ing

HP

B b

reak

s•

Each

alte

rnat

ive

conf

igur

atio

n co

nsid

ered

can

be

com

bine

d w

ith a

llev

alua

ted

embe

dmen

t alte

rnat

ives

•D

esig

n of

the

reac

tor a

nd H

PB

pre

ssur

e bo

unda

ry a

nd p

assi

ve s

afet

ych

arac

teris

tics

will

limit

the

pote

ntia

l for

air

ingr

ess;

cap

abilit

y to

sup

port

ad

hoc

emer

genc

y ac

tions

to in

ert t

he re

acto

r cor

e ca

vity

via

inje

ctio

n of

he

lium

or n

itrog

en is

inhe

rent

in th

e P

BM

R fu

el h

andl

ing

syst

em d

esig

n•

All

desi

gns

cons

ider

ed li

mit

the

supp

ly o

f out

side

air

by c

ontro

lling

build

ing

leak

rate

but

leak

rate

var

ies

amon

g th

e co

nsid

ered

alte

rnat

ives

Slid

e30

PB

MR

TE

AM

Fiss

ion

Prod

uct T

rans

port

and

Pre

ssur

e R

elie

f Pa

thw

ays

durin

g H

PB L

eak/

Bre

ak E

vent

s

•H

BP le

aks/

brea

ks in

Rea

ctor

Cav

ity–

Exp

ands

to re

acto

r cav

ity v

ia e

ngin

eere

d flo

w p

ath

to IH

X c

avity

–Fl

ows

thro

ugh

engi

neer

ed fl

ow p

ath

from

IHX

cav

ity to

PR

S v

ent s

haft

•H

PB le

ak/b

reak

in C

CS

Cav

ity (n

ot s

how

n in

follo

win

g sc

hem

atic

s)–

Exp

ands

to re

acto

r cav

ity v

ia e

ngin

eere

d flo

w p

ath

to IH

X c

avity

–Fl

ows

thro

ugh

engi

neer

ed fl

ow p

ath

from

IHX

cav

ity to

PR

S v

ent s

haft

•H

PB le

ak/b

reak

in F

HSS

or H

SS a

reas

–E

xpan

ds to

FH

SS

or H

SS

spa

ces

–Fl

ows

thro

ugh

engi

neer

ed fl

ow p

ath

to P

RS

ven

t sha

ft•

HP

B le

ak/b

reak

in IH

X s

pace

on

eith

er th

e P

HTS

or t

he S

HTS

sid

e–

Exp

ands

to IH

X s

pace

–Fl

ows

thro

ugh

engi

neer

ed fl

ow p

ath

to P

RS

ven

t sha

ft

Som

e ex

pans

ion

into

oth

er b

uild

ing

spac

es m

ay a

lso

occu

r thr

ough

impe

rfect

doo

rs, s

eals

, and

pen

etra

tions

. Le

akag

e fro

m R

B e

nvel

ope

to e

nviro

nmen

t will

depe

nd o

n en

velo

pe le

ak ra

te.

Slid

e31

PB

MR

TE

AM

Alte

rnat

ive

1aU

nfilt

ered

and

Ven

ted

Rea

ctor

Cav

ity

IHX

Cav

ity

All

Oth

er C

ompa

rtm

ents

with

PH

TS H

PB S

SCs

incl

udin

g H

SS a

nd F

HSS

Pres

sure

Rel

ief

Path

way

Prim

ary

HTS

Seco

ndar

yH

TS

Slid

e32

PB

MR

TE

AM

NH

SB A

ltern

ativ

e 1a

Unf

ilter

ed a

nd V

ente

d

Key

Des

ign

Feat

ures

:•

Pre

ssur

ized

Zon

e Le

ak R

ate

-50-

100

vol %

/day

(pos

t blo

w-d

own)

–D

oors

and

pen

etra

tions

-Ind

ustri

al g

rade

–Se

alin

g an

d fin

ish

–in

dust

rial g

rade

•Pr

essu

re R

elie

f Pat

hway

–Is

olat

ed fr

om n

orm

al H

VAC

by

low

pre

ssur

e ru

ptur

e pa

nels

–C

lean

sur

face

s to

min

imiz

e de

posi

tion

durin

g bl

ow-d

own

–Sm

ooth

tran

sitio

ns fr

om ro

om to

room

to m

inim

ize

flow

resi

stan

ce–

Inte

rnal

ope

n ve

nts

to li

mit

com

partm

ent o

ver-p

ress

ure

and

cont

rol r

elea

se p

athw

ay–

Exte

rnal

ope

n ve

nts

to a

relie

f sha

ft–

No

capa

bilit

y fo

r ven

t sha

ft re

-clo

sure

•Fi

ltrat

ion

of A

ccid

enta

l rel

ease

s–

No

filtra

tion

of b

low

-dow

n ph

ase

of re

leas

e. S

ome

pass

ive

deco

ntam

inat

ion

due

to p

late

-ou

t, co

nden

satio

n, a

nd d

ecay

.–

Pass

ive

plat

e-ou

t, co

nden

satio

n, s

ettli

ng, r

adio

activ

e de

cay

for d

elay

ed fu

el re

leas

e•

Nor

mal

HVA

C is

olat

ed fr

om p

ress

ure

trans

ient

by

isol

atio

n de

vice

but h

as

no s

afet

y fu

nctio

n•

Pos

t-eve

nt c

lean

up u

sing

nor

mal

HV

AC

if a

vaila

ble

Slid

e33

PB

MR

TE

AM

Pros

and

Con

s of

Alte

rnat

ive

1a

•M

inim

izes

pre

ssur

e tra

nsie

nt•

Inhe

rent

ly re

liabl

e as

it is

co

mpl

etel

y pa

ssiv

e

Adv

anta

ge

•M

ore

diffi

cult

to c

ontro

l re

leas

e of

air

activ

atio

n pr

oduc

ts d

urin

g no

rmal

op

erat

ion

•N

o en

gine

ered

feat

ures

to

reta

in a

ccid

ent r

elat

ed fi

ssio

n pr

oduc

ts o

r lim

it ai

r ing

ress

to

bui

ldin

g fo

llow

ing

blow

-do

wn

•M

ay re

quire

ver

y la

rge

HV

AC

equ

ipm

ent t

o m

aint

ain

zoni

ng a

nd z

one

pres

sure

gr

adie

nts

Ope

n ve

nt p

aths

Dis

adva

ntag

eFe

atur

e

Slid

e34

PB

MR

TE

AM

Alte

rnat

ive

1bU

nfilt

ered

and

Ven

ted

with

Blo

wou

t Pan

els

Rea

ctor

Cav

ity

IHX

Cav

ity

All

Oth

er C

ompa

rtm

ents

with

PH

TS H

PB S

SCs

incl

udin

g H

SS a

nd F

HSS

Pres

sure

Rel

ief

Path

way

Prim

ary

HTS

Seco

ndar

yH

TS

Rup

ture

Pan

el -

norm

ally

clo

sed

- op

ens

with

inte

rnal

pre

ssur

e -

does

not

re-c

lose

Slid

e35

PB

MR

TE

AM

NH

SB A

ltern

ativ

e 1b

Unf

ilter

ed a

nd V

ente

d w

ith B

low

out P

anel

s K

ey D

esig

n Fe

atur

es:

•B

uild

ing

Leak

Rat

e-5

0-10

0 vo

l % /d

ay (p

ost b

low

-dow

n)–

Doo

rs a

nd p

enet

ratio

ns-I

ndus

trial

gra

de–

Sea

ling

and

finis

h –

indu

stria

l gra

de•

Pres

sure

Rel

ief P

athw

ay–

Isol

ated

from

nor

mal

HV

AC

by

low

pre

ssur

e ru

ptur

e pa

nels

–C

lean

sur

face

s to

min

imiz

e de

posi

tion

durin

g bl

ow-d

own

–S

moo

th tr

ansi

tions

from

room

to ro

om to

min

imiz

e flo

w re

sist

ance

–In

tern

al b

low

out p

anel

s to

lim

it co

mpa

rtmen

t ove

r-pr

essu

re a

nd c

ontro

l rel

ease

pa

thw

ay–

Ext

erna

l blo

wou

t pan

els

to c

ontro

l rel

ease

pat

h vi

a a

relie

f sha

ft–

No

capa

bilit

y fo

r ven

t sha

ft re

-clo

sure

•Fi

ltrat

ion

of A

ccid

enta

l rel

ease

s–

No

filtra

tion

of b

low

-dow

n ph

ase

of re

leas

e. S

ome

pass

ive

deco

ntam

inat

ion

due

topl

ateo

ut, c

onde

nsat

ion,

and

dec

ay.

–P

assi

ve p

late

-out

, con

dens

atio

n, s

ettli

ng, r

adio

activ

e de

cay

for d

elay

ed fu

el

rele

ase

•N

orm

al H

VAC

isol

ated

from

pre

ssur

e tra

nsie

nt b

y is

olat

ion

devi

cebu

t has

no

saf

ety

func

tion

•P

ost-e

vent

cle

anup

usi

ng n

orm

al H

VA

C if

ava

ilabl

e

Slid

e36

PB

MR

TE

AM

Pros

and

Con

s of

Alte

rnat

ive

1b

•Eas

ier t

o co

ntro

l rel

ease

of

air a

ctiv

atio

n pr

oduc

ts d

urin

g no

rmal

ope

ratio

n•E

asie

r for

HV

AC

to c

ontro

l pr

essu

re z

ones

Adv

anta

ge

•N

o en

gine

ered

feat

ures

to

reta

in a

ccid

ent r

elat

ed fi

ssio

n pr

oduc

ts

Ven

t pat

hs re

mai

n op

en

afte

r rup

ture

•Inc

reas

es p

ress

ure

trans

ient

and

des

ign

pres

sure

•Rel

iabl

e bu

t les

s so

than

op

en v

ent p

ath

Rup

ture

pan

el o

n ve

nt

path

s

Dis

adva

ntag

eFe

atur

e

Slid

e37

PB

MR

TE

AM

Alte

rnat

ive

2Pa

rtia

lly F

ilter

ed a

nd V

ente

d w

ith B

low

out p

anel

s

Rea

ctor

Cav

ity

IHX

Cav

ity

All

Oth

er C

ompa

rtm

ents

with

PH

TS H

PB S

SCs

incl

udin

g H

SS a

nd F

HSS

Pres

sure

Rel

ief

Path

way

Prim

ary

HTS

Seco

ndar

yH

TS

Rup

ture

Pan

el -

norm

ally

clo

sed

- op

ens

with

inte

rnal

pre

ssur

e -

does

not

re-c

lose

Rec

losa

ble

Dam

per

Filte

r

Slid

e38

PB

MR

TE

AM

NH

SB A

ltern

ativ

e 2

Del

ayed

Rel

ease

Filt

ered

and

Ven

ted

with

Blo

wou

t Pan

els

Key

Des

ign

Feat

ures

:•

Bui

ldin

g Le

ak R

ate

-25-

50 v

ol %

/day

(Ass

umed

dur

ing

dela

yed

rele

ase)

–D

oors

and

pen

etra

tions

–so

ft an

d in

flata

ble

seal

s–

Sea

ling

and

finis

h –

Bui

ldin

g bo

unda

ry s

eale

d at

ele

men

t joi

nts

by g

aske

ts,

caul

king

, or w

eldi

ng•

Pres

sure

Rel

ief P

athw

ay–

Isol

ated

from

nor

mal

HV

AC

by

low

pre

ssur

e ru

ptur

e pa

nels

–C

lean

sur

face

s to

min

imiz

e de

posi

tion

durin

g bl

ow-d

own

–S

moo

th tr

ansi

tions

from

room

to ro

om to

min

imiz

e flo

w re

sist

ance

–In

tern

al b

low

out p

anel

s to

lim

it co

mpa

rtmen

t ove

rpre

ssur

e an

d co

ntro

l rel

ease

pa

thw

ay–

Ext

erna

l blo

wou

t pan

els

to c

ontro

l rel

ease

pat

h vi

a re

lief s

haft

–E

xit p

oint

thro

ugh

pres

sure

act

uate

d da

mpe

r dur

ing

initi

al b

low

-dow

n an

d th

roug

h P

ost-E

vent

HV

AC

filtr

atio

n fo

llow

ing

pres

sure

equ

aliz

atio

n an

d da

mpe

rre

-clo

sure

–H

VA

C fi

lter o

ptio

ns in

clud

e si

nter

ed m

etal

, HE

PA

, wet

scr

ubbe

rs,a

nd o

ther

s •

Nor

mal

HV

AC

isol

ated

from

pre

ssur

e tra

nsie

nt b

y bl

ast p

anel

s -

HV

AC

rest

art a

fter p

ress

ure

equa

lizat

ion

to a

ssis

t pos

t blo

w-d

own

filtra

tion

Slid

e39

PB

MR

TE

AM

Pros

and

Con

s of

Alte

rnat

ive

2

•R

educ

es th

e de

laye

d fu

el

rele

ase

sour

ce te

rm

•M

inim

izes

pre

ssur

e tra

nsie

nt b

y op

enin

g ve

nt

durin

g la

rge

HP

B b

reak

•In

crea

ses

effe

ctiv

enes

s of

fil

ter

•E

asie

r to

cont

rol r

elea

se o

f ai

r act

ivat

ion

prod

ucts

dur

ing

norm

al o

pera

tion

•E

asie

r for

HV

AC

to c

ontro

l pr

essu

re z

ones

Adv

anta

ge

•Inc

reas

es c

ost

Filte

r on

vent

exi

t

•Inc

reas

es c

ost

Re-

clos

eabl

e da

mpe

r on

vent

exi

t

•Inc

reas

es p

ress

ure

trans

ient

Rup

ture

pan

el o

n ve

nt

path

s

Dis

adva

ntag

eFe

atur

e

Slid

e40

PB

MR

TE

AM

Rea

ctor

Bui

ldin

g A

ltern

ativ

e 3a

Filte

red

and

Vent

ed w

ith B

low

out P

anel

s

Rea

ctor

Cav

ity

IHX

Cav

ity

All

Oth

er C

ompa

rtm

ents

with

PH

TS H

PB S

SCs

incl

udin

g H

SS a

nd F

HSS

Pres

sure

Rel

ief

Path

way

Prim

ary

HTS

Seco

ndar

yH

TS

Rup

ture

Pan

el -

norm

ally

clo

sed

- op

ens

with

inte

rnal

pre

ssur

e -

does

not

re-c

lose

Filte

r

Slid

e41

PB

MR

TE

AM

NH

SB A

ltern

ativ

e 3a

Filte

red

and

Vent

ed w

ith B

low

out P

anel

s

Key

Des

ign

Feat

ures

:B

uild

ing

Leak

Rat

e-2

5-50

vol

% /d

ay (d

urin

g de

laye

d re

leas

e)–

Doo

rs a

nd p

enet

ratio

ns–

soft

and

infla

tabl

e se

als

–S

ealin

g an

d fin

ish

–B

uild

ing

boun

dary

sea

led

at e

lem

ent j

oint

s by

gas

kets

, ca

ulki

ng, o

r wel

ding

•Pr

essu

re R

elie

f Pat

hway

–Is

olat

ed fr

om n

orm

al H

VA

C b

y lo

w p

ress

ure

rupt

ure

pane

ls–

Rou

gh s

urfa

ces

to m

axim

ize

depo

sitio

n du

ring

blow

-dow

n–

Sha

rp tr

ansi

tions

from

room

to ro

om to

max

imiz

e flo

w re

sist

ance

–In

tern

al b

low

out p

anel

s to

lim

it co

mpa

rtmen

t ove

rpre

ssur

e an

d co

ntro

l rel

ease

pa

thw

ay–

Ext

erna

l blo

wou

t pan

els

to c

ontro

l rel

ease

pat

h vi

a re

lief s

haft

–E

xit p

oint

via

pas

sive

filte

rs d

urin

g bl

ow-d

own

–pa

ssiv

e fil

ter m

ust t

oler

ate

pres

sure

and

tem

pera

ture

–E

xit p

oint

thro

ugh

filte

rfol

low

ing

pres

sure

equ

aliz

atio

n–

Pos

t blo

w-d

own

filte

r opt

ions

incl

ude

sint

ered

met

al, H

EP

A, w

et s

crub

bers

,and

othe

rs•

Nor

mal

HVA

C is

olat

ed fr

om p

ress

ure

trans

ient

by

blas

t pan

els

Slid

e42

PB

MR

TE

AM

Pros

and

Con

s of

Alte

rnat

ive

3a

•Red

uces

rele

ase

of ra

dion

uclid

es

thro

ugh

build

ing

boun

dary

(filt

er

bypa

ss)

•R

educ

es th

e pr

ompt

re

leas

e so

urce

term

•R

educ

es th

e de

laye

d fu

el re

leas

e so

urce

term

•Eas

ier t

o co

ntro

l rel

ease

of a

ir ac

tivat

ion

prod

ucts

dur

ing

norm

al

oper

atio

n•

Eas

ier f

or H

VA

C to

con

trol p

ress

ure

zone

s

Adv

anta

ge

•In

crea

ses

cost

Red

uced

leak

rate

for

pres

suriz

ed z

one

•In

crea

ses

pres

sure

tran

sien

tFi

lter o

n ve

nt e

xit

•In

crea

ses

pres

sure

tran

sien

tR

uptu

re p

anel

on

vent

pa

ths

Dis

adva

ntag

eFe

atur

e

Slid

e43

PB

MR

TE

AM

Alte

rnat

ive

3bFi

ltere

d an

d Ve

nted

with

Blo

wou

t Pan

els

and

Expa

nsio

n Vo

lum

e

Rea

ctor

Cav

ity

IHX

Cav

ity

All

Oth

er C

ompa

rtm

ents

with

PH

TS H

PB S

SCs

incl

udin

g H

SS a

nd F

HSS

Pres

sure

Rel

ief

Path

way

Prim

ary

HTS

Seco

ndar

yH

TS

Rup

ture

Pan

el -

norm

ally

clo

sed

- op

ens

with

inte

rnal

pre

ssur

e -

does

not

re-c

lose

Filte

r

Expa

nsio

nVo

lum

e

Slid

e44

PB

MR

TE

AM

Alte

rnat

ive

3bFi

ltere

d an

d Ve

nted

with

Blo

wou

t Pan

els

and

Expa

nsio

n Vo

lum

eK

ey D

esig

n Fe

atur

es:

Bui

ldin

g Le

ak R

ate

-25-

50 v

ol %

/day

(dur

ing

dela

yed

fuel

rele

ase)

–D

oors

and

pen

etra

tions

-Nuc

lear

Gra

de –

soft

and

infla

tabl

e se

als

–S

ealin

g an

d fin

ish

–B

uild

ing

boun

dary

sea

led

at e

lem

ent j

oint

s by

gas

kets

, ca

ulki

ng, o

r wel

ding

•Pr

essu

re R

elie

f Pat

hway

–Is

olat

ed fr

om n

orm

al H

VA

C b

y lo

w p

ress

ure

rupt

ure

pane

ls–

Rou

gh s

urfa

ces

to m

axim

ize

depo

sitio

n du

ring

blow

dow

n–

Sha

rp tr

ansi

tions

from

room

to ro

om to

max

imiz

e flo

w re

sist

ance

–In

tern

al b

low

out p

anel

s an

d ex

pans

ion

volu

me

to li

mit

com

partm

ent

over

pres

sure

and

con

trol r

elea

se p

athw

ay–

Ext

erna

l blo

wou

t pan

els

to c

ontro

l rel

ease

pat

h vi

a re

lief s

haft

–E

xit p

oint

via

pas

sive

filte

rs d

urin

g bl

ow-d

own

–pa

ssiv

e fil

ter m

ust t

oler

ate

pres

sure

and

tem

pera

ture

(sin

tere

d m

etal

?)–

Exi

t poi

nt th

roug

h fil

ter f

ollo

win

g pr

essu

re e

qual

izat

ion

–P

ost b

low

-dow

n fil

ter o

ptio

ns in

clud

e si

nter

ed m

etal

, HE

PA

, wet

scr

ubbe

rs,a

ndot

hers

•N

orm

al H

VAC

isol

ated

from

pre

ssur

e tra

nsie

nt b

y bl

ast p

anel

s

Slid

e45

PB

MR

TE

AM

Pros

and

Con

s of

Alte

rnat

ive

3b

•Incr

ease

s pr

essu

re

trans

ient

•Ret

ains

som

e ra

dion

uclid

es fr

om

prom

pt re

leas

e•R

etai

ns s

ome

radi

onuc

lides

from

de

laye

d re

leas

e

Filte

r on

vent

exi

t

•Red

uces

rele

ase

of ra

dion

uclid

es

thro

ugh

build

ing

boun

dary

(filt

er

bypa

ss)

•Red

uces

pre

ssur

e tra

nsie

nt

•Con

trols

rele

ase

of a

ir ac

tivat

ion

prod

ucts

Adv

anta

ge

•Incr

ease

s co

stR

educ

ed le

ak ra

te fo

r pr

essu

rized

zon

e

•Incr

ease

s co

stIn

crea

sed

expa

nsio

n vo

lum

e

•Incr

ease

s pr

essu

re

trans

ient

Rup

ture

pan

el o

n ve

nt

path

s

Dis

adva

ntag

eFe

atur

e

Slid

e46

PB

MR

TE

AM

Alte

rnat

ive

4aPr

essu

re R

etai

ning

with

Inte

rnal

Blo

wou

t Pan

els

Rea

ctor

Cav

ity

IHX

Cav

ity

All

Oth

er C

ompa

rtm

ents

with

PH

TS H

PB S

SCs

incl

udin

g H

SS a

nd F

HSSPr

imar

yH

TSSe

cond

ary

HTS

Rup

ture

Pan

el -

norm

ally

clo

sed

- op

ens

with

inte

rnal

pre

ssur

e -

does

not

re-c

lose

Slid

e47

PB

MR

TE

AM

Alte

rnat

ive

4aPr

essu

re R

etai

ning

with

Inte

rnal

Blo

wou

t Pan

els

Key

Des

ign

Feat

ures

:•

Bui

ldin

g Le

ak R

ate

-0.1

-1 v

ol %

/day

–D

oors

and

pen

etra

tions

-Airl

ock

door

s w

ith in

flata

ble

seal

s, h

ard

pene

tratio

ns (w

elde

d)–

Sea

ling

and

finis

h –

Epo

xy o

r met

al b

uild

ing

liner

, wel

ded

–B

uild

ing

geom

etry

may

cha

nge

to e

nabl

e re

sist

ance

to p

ress

ure

load

s–

Achi

evin

g 1%

/d le

ak ra

te m

ay c

ompl

icat

e de

sign

for R

CC

S•

Pres

sure

Rel

ief P

athw

ay–

Inte

rnal

blo

wou

t pan

els

to li

mit

com

partm

ent o

verp

ress

ure

and

cont

rol r

elea

se p

athw

ay–

No

exte

rnal

rele

ase

poin

t oth

er th

an n

orm

al le

akag

e an

d po

st

even

t pur

ge li

nes

•N

orm

al H

VAC

isol

ated

from

pre

ssur

e tra

nsie

nt b

y bl

ast p

anel

s•

Pos

t-eve

nt c

lean

up u

sing

ded

icat

ed e

xtra

ctio

n fa

ns a

nd fi

lters

and

pe

rhap

s he

at e

xcha

nger

s (to

pro

tect

filte

rs)

–P

ost e

vent

bui

ldin

g de

pres

suriz

atio

n w

ill in

clud

e lo

ng li

ved

nobl

ega

ses

Slid

e48

PB

MR

TE

AM

Pros

and

Con

s of

Alte

rnat

ive

4a

•In

crea

ses

pres

sure

tran

sien

t•

Incr

ease

s co

st•

May

requ

ire c

urve

d w

alls

and

sl

abs

to re

sist

pre

ssur

e tra

nsie

nt

•A

ll P

HTS

HP

B re

leas

es

reta

ined

with

in b

uild

ing

Elim

inat

ion

of e

ngin

eere

d ve

nt p

ath

•R

educ

es re

leas

e of

ra

dion

uclid

es th

roug

h bu

ildin

g bo

unda

ry

Adv

anta

ge

•Req

uire

s en

gine

ered

pe

netra

tions

•May

requ

ire d

esig

n ch

ange

s in

RC

CS,

HVA

C•I

ncre

ased

requ

irem

ents

for

leak

rate

test

ing

•Con

tain

men

t fai

lure

s m

odes

m

ay re

duce

effe

ctiv

enes

s

Red

uced

leak

rate

for

pres

suriz

ed z

one

Dis

adva

ntag

eFe

atur

e

Slid

e49

PB

MR

TE

AM

Alte

rnat

ive

4bPr

essu

re R

etai

ning

with

Inte

rnal

Blo

wou

t Pan

els

and

Expa

nsio

n Vo

lum

e

Prim

ary

HTS

Seco

ndar

yH

TS

Rup

ture

Pan

el -

norm

ally

clo

sed

- op

ens

with

inte

rnal

pre

ssur

e -

does

not

re-c

lose

Expa

nsio

nVo

lum

eR

eact

or C

avity

IHX

Cav

ity

All

Oth

er C

ompa

rtm

ents

with

PH

TS H

PB S

SCs

incl

udin

g H

SS a

nd F

HSS

Slid

e50

PB

MR

TE

AM

Alte

rnat

ive

4bPr

essu

re R

etai

ning

with

Inte

rnal

B

low

out P

anel

s an

d Ex

pans

ion

Volu

me

Key

Des

ign

Feat

ures

:•

Bui

ldin

g Le

ak R

ate

-0.1

-1 v

ol %

/day

–D

oors

and

pen

etra

tions

-Airl

ock

door

s w

ith in

flata

ble

seal

s, h

ard

pene

tratio

ns (w

elde

d)–

Sea

ling

and

finis

h –

Epo

xy o

r met

al b

uild

ing

liner

, wel

ded

–B

uild

ing

geom

etry

may

cha

nge

to e

nabl

e re

sist

ance

to p

ress

ure

load

s–

Achi

evin

g 1%

/d le

ak ra

te m

ay c

ompl

icat

e de

sign

for R

CC

S•

Pres

sure

Rel

ief P

athw

ay–

Inte

rnal

blo

wou

t pan

els

and

expa

nsio

n vo

lum

e to

lim

it co

mpa

rtmen

tov

erpr

essu

re a

nd c

ontro

l rel

ease

pat

hway

–N

o ex

tern

al re

leas

e po

int o

ther

than

nor

mal

leak

age

and

post

eve

ntpu

rge

lines

•N

orm

al H

VAC

isol

ated

from

pre

ssur

e tra

nsie

nt b

y bl

ast p

anel

s•

Pos

t-eve

nt c

lean

up u

sing

ded

icat

ed e

xtra

ctio

n fa

ns a

nd fi

lters

and

pe

rhap

s he

at e

xcha

nger

s (to

pro

tect

filte

rs)

–P

ost e

vent

bui

ldin

g de

pres

suriz

atio

n w

ill in

clud

e lo

ng li

ved

nobl

ega

ses

Slid

e51

PB

MR

TE

AM

Pros

and

Con

s of

Alte

rnat

ive

4b

•Incr

ease

s co

st•D

ecre

ases

pre

ssur

e tra

nsie

ntIn

crea

sed

Exp

ansi

on

Vol

ume

•Incr

ease

s pr

essu

re tr

ansi

ent

•Incr

ease

s co

st•M

ay re

quire

cur

ved

wal

ls a

nd

slab

s to

resi

st p

ress

ure

trans

ient

•All

PH

TS H

PB

rele

ases

re

tain

ed w

ithin

bui

ldin

gE

limin

atio

n of

eng

inee

red

vent

pat

h

•Red

uces

rele

ase

of

radi

onuc

lides

thro

ugh

build

ing

boun

dary

Adv

anta

ge

•Req

uire

s en

gine

ered

pe

netra

tions

•May

requ

ire d

esig

n ch

ange

s in

R

CC

S, H

VA

C•I

ncre

ased

requ

irem

ents

for l

eak

rate

test

ing

•Con

tain

men

t fai

lure

s m

odes

m

ay re

duce

effe

ctiv

enes

s

Red

uced

leak

rate

for

pres

suriz

ed z

one

Dis

adva

ntag

eFe

atur

e

Slid

e52

PB

MR

TE

AM

Sum

mar

y of

Alte

rnat

ive

RB

Con

figur

atio

ns

Rad

ionu

clid

e Fi

ltrat

ion

No.

Des

ign

Des

crip

tion

Pres

suriz

edZo

ne L

eak

Rat

eV

ol %

/day

NH

SBB

ound

ary

Leak

Rat

e V

ol%

/day

Pres

sure

Rel

ief D

esig

n Fe

atur

es

Post

blo

w-

dow

n re

-cl

osur

e of

PR

S sh

aft?

Blo

w-

dow

nph

ase

Del

ayed

fuel

rele

ase

phas

e

1a.

Unf

ilter

ed a

nd v

ente

d50

-100

50-1

00O

pen

vent

No

Non

ePa

ssiv

e

1bU

nfilt

ered

and

ven

ted

with

blo

wou

t pan

els

50-1

0050

-100

Inte

rnal

+ E

xter

nal b

low

out

pane

lsN

oN

one

Pass

ive

2Pa

rtial

ly fi

ltere

d an

d ve

nted

with

blo

wou

t pa

nels

25-5

050

-100

Inte

rnal

+ E

xter

nal b

low

out

pane

lsY

esN

one

Act

ive

HV

AC

3aFi

ltere

d an

d ve

nted

with

bl

owou

t pan

els

25-5

050

-100

Inte

rnal

+ E

xter

nal b

low

out

pane

lsY

esPa

ssiv

eA

ctiv

eH

VA

C

3bFi

ltere

d an

d ve

nted

with

bl

owou

t pan

els a

nd

expa

nsio

n vo

lum

e

25-5

050

-100

Inte

rnal

+ E

xter

nal b

low

out

pane

ls +

exp

ansi

on v

olum

eY

esPa

ssiv

eA

ctiv

eH

VA

C

4aPr

essu

re re

tain

ing

with

in

tern

al b

low

out p

anel

s0.

1-1

50-1

00In

tern

al b

low

out p

anel

sN

/APa

ssiv

ePa

ssiv

e

4bPr

essu

re re

tain

ing

with

in

tern

al b

low

out p

anel

s an

d ex

pans

ion

volu

me

0.1-

150

-100

Inte

rnal

blo

wou

t pan

els +

ex

pans

ion

volu

me

N/A

Pass

ive

Pass

ive

Slid

e53

PB

MR

TE

AM

Som

e C

ause

and

Effe

ct R

elat

ions

hips

•D

ecre

ase

in v

ent f

low

vol

ume

caus

es:

–In

crea

se in

pea

k tra

nsie

nt p

ress

ure

•D

ecre

ase

in b

uild

ing

leak

rate

cau

ses:

–In

crea

se in

pea

k tra

nsie

nt p

ress

ure

–In

crea

sed

soph

istic

atio

n in

pen

etra

tions

–In

crea

sed

diffi

culty

in d

esig

ning

sys

tem

s th

at o

pera

te a

cros

s bu

ildin

gbo

unda

ry (R

CC

S, S

HTS

, HVA

C)

•In

crea

se in

pea

k tra

nsie

nt p

ress

ure

caus

es:

–In

crea

sed

load

s on

wal

ls a

nd s

labs

•In

crea

se in

wal

l and

sla

b lo

ads

caus

es:

–In

crea

se in

wal

l and

sla

b th

ickn

ess,

or

–C

hang

e to

cur

ved

wal

ls a

nd s

labs

(rou

nd b

uild

ing)

Slid

e54

PB

MR

TE

AM

Sum

mar

y of

Rea

ctor

Bui

ldin

g Fu

nctio

nal

Req

uire

men

ts fo

r Rad

ionu

clid

e R

eten

tion

and

Pres

sure

Cap

acity

Slid

e55

PB

MR

TE

AM

Obj

ectiv

es

•E

stab

lish

requ

irem

ents

for t

he N

GN

P R

eact

or B

uild

ing

with

re

spec

t to:

–R

eten

tion

of ra

dion

uclid

es d

urin

g se

lect

ed li

cens

ing

basi

s ev

ents

(LB

Es)

invo

lvin

g H

PB

leak

s an

d br

eaks

–P

ress

ure

capa

city

of t

he re

acto

r bui

ldin

g fo

r LB

Es

–D

imen

sion

s of

the

Rea

ctor

Bui

ldin

g Ve

nt V

olum

e (R

BVV)

–Le

ak R

ate

of th

e R

BVV

–P

ress

ure

set-p

oint

s of

rupt

ure

pane

ls a

nd H

VA

C d

ampe

rs

•E

valu

ate

iden

tifie

d R

eact

or B

uild

ing

conc

epts

with

resp

ect t

o ab

ove

requ

irem

ents

•E

stab

lish

radi

olog

ical

Saf

ety

Mar

gins

of i

dent

ified

Rea

ctor

B

uild

ing

conc

epts

aga

inst

Top

Lev

el R

egul

ator

y C

riter

ia•

Eva

luat

e ph

enom

ena

resp

onsi

ble

for t

rend

s in

rele

ases

ove

r a

wid

e se

t of a

naly

sis

case

s

Slid

e56

PB

MR

TE

AM

Bou

ndar

y C

ondi

tions

•To

p Le

vel R

egul

ator

y C

riter

ia–

Site

Bou

ndar

y D

ose

Ass

essm

ent

•To

talE

ffect

ive

Dos

e E

quiv

alen

t (TE

DE

) = E

ffect

ive

Dos

e E

quiv

alen

t (fo

r ext

erna

l exp

osur

es) p

lus

Com

mitt

ed

Effe

ctiv

e D

ose

Equ

ival

ent (

for i

nter

nal e

xpos

ures

).•

EP

A P

rote

ctiv

e A

ctio

n G

uide

s (P

AG

s)–

PA

G fo

r TE

DE

: 1

rem

/eve

nt–

PA

G fo

r Thy

roid

Dos

e:

5 re

mpe

r eve

nt•

10 C

FR 5

0.34

Crit

eria

for D

esig

n B

asis

Acc

iden

ts–

TED

E:

2

5 re

m/e

vent

–Th

yroi

d:

75

rem

/eve

nt

Slid

e57

PB

MR

TE

AM

Con

stra

ints

•N

o in

tegr

ated

ana

lysi

s m

odel

ava

ilabl

e fo

r Blo

wdo

wn

Pha

se,

Hea

tup

Pha

se a

nd C

oold

own

Pha

se c

oupl

ed w

ith a

Rea

ctor

B

uild

ing

and

Offs

ite D

ose

Mod

el•

Pro

ject

sch

edul

e an

d bu

dget

con

stra

ints

•S

elec

ted

LBE

slim

ited

to D

epre

ssur

ized

Los

s of

For

ced

Coo

ling

(DLO

FC) c

ases

for a

rang

e of

bre

ak s

izes

in th

e C

ore

Inle

t Pip

e–

Equi

vale

nt b

reak

siz

es o

f 2,3

,10,

100,

and

100

0mm

(DEG

B of

the

710m

m C

IP)

•Fo

r est

ablis

hing

RB

des

ign

pres

sure

a d

esig

n ba

sis

even

t inv

olvi

ng10

00m

m D

EG

B o

f the

SH

TS H

PB

is a

ssum

ed–

Lack

of a

ctiv

e co

olin

g fo

r SH

TS b

ased

on

PC

DR

des

ign

intro

duce

s la

rge

unce

rtain

ties

in re

liabi

lity

of h

ot g

as p

ipe

liner

and

insu

latio

nw

hose

failu

re w

ould

dire

ctly

lead

to m

ajor

rupt

ure

of th

e H

PB

Slid

e58

PB

MR

TE

AM

Rad

ionu

clid

e In

form

atio

n fo

r DLO

FC C

ases

•R

adio

nucl

ide

Inve

ntor

y in

HP

B a

t Beg

inni

ng o

f In

cide

nt–

Circ

ulat

ing

Act

ivity

at E

nd o

f Life

Ful

l Pow

er

Ope

ratio

n–

Life

time

Pla

teou

t and

Dep

ositi

on A

ctiv

ity–

Equ

ilibr

ium

Dus

t Act

ivity

–S

ourc

e: N

GN

P C

onta

min

atio

n C

ontro

l Rep

ort

(050

108)

Slid

e59

PB

MR

TE

AM

Rad

ionu

clid

e In

form

atio

n fo

r DLO

FC C

ase

(con

tinue

d)

•Tr

ansi

ent R

adio

nucl

ide

Rel

ease

from

Fue

l D

urin

g In

cide

nt–

Cor

e te

mpe

ratu

re tr

ansi

ent d

urin

g D

LOFC

–C

alcu

latio

ns a

ssum

e at

mos

pher

ic p

ress

ure

at t

= 0

–Ti

me

depe

nden

t act

ivity

rele

ase

durin

g D

LOFC

cor

e he

atup

as

a fu

nctio

n of

cor

e te

mpe

ratu

re fo

r dur

atio

n of

rele

ases

from

the

core

–S

ourc

e: E

xist

ing

PB

MR

cal

cula

tions

for a

500

MW

t de

sign

with

diff

eren

t ope

ratin

g co

nditi

ons

Slid

e60

PB

MR

TE

AM

Cor

e Fu

el T

ime

at T

empe

ratu

re D

istr

ibut

ion

durin

g D

LOFC

PB V

olum

e D

istr

ibut

ion

duri

ng D

LOFC

0102030405060708090100

040

8012

016

020

0

Tim

e (h

r)

% PB Volume

>300

°C

>400

°C

>500

°C

>600

°C>7

00°C

>800

°C

>900

°C

>100

0 °C

>110

>120

0 °C

>130

0 °C

>140

0 °C

>150

0 °C

>160

0 °C

>170

0°C

>180

0°C

Slid

e61

PB

MR

TE

AM

Act

ivity

Rel

ease

from

Fue

l dur

ing

DLO

FC

DLO

FC R

elea

se fr

om F

uel -

500

MW

PH

OS

1.0E

+16

1.0E

+17

1.0E

+18

1.0E

+19

1.0E

+20

1.0E

+21

1.0E

+22

025

5075

100

125

150

175

200

225

250

275

300

Hou

rs

Total released inventory (atoms)

Cs

-137

Ag-1

11I-1

31Sr

-90

Slid

e62

PB

MR

TE

AM

Tech

nica

l App

roac

h

•U

se te

chni

cal i

nfor

mat

ion

avai

labl

e to

dev

elop

a

sim

plifi

ed 3

vol

ume

inte

grat

ed m

odel

of P

HTS

, RB

and

S

ite B

ound

ary

Dos

e–

Vol

ume

1:

PH

TS H

ot V

olum

e–

Vol

ume

2:

PH

TS C

old

Vol

ume

with

Bre

ak–

Vol

ume

3:

Gen

eral

ized

RB

Ven

t Vol

ume

(RB

VV

)

•C

onst

ant V

olum

e S

ize

with

sen

sitiv

ity a

naly

sis

of

alte

rnat

ive

RB

VV

dim

ensi

ons

•D

urin

g he

atup

the

expa

ndin

g ho

t vol

ume

trans

fers

hot

co

olan

t mas

s &

ene

rgy

to c

old

volu

me,

mix

es in

stan

tly

in c

old

volu

me.

Slid

e63

PB

MR

TE

AM

Tech

nica

l App

roac

h (c

ontin

ued)

•R

BV

V m

odel

cap

able

of s

imul

atin

g al

l pro

pose

d R

B d

esig

n al

tern

ativ

es.

•To

tal v

olum

e of

RB

incl

udin

g ve

nted

and

non

-ven

ted

area

s is

ap

prox

imat

ely

100,

000m

3pe

r PC

DR

•V

olum

e of

RB

VV

of 1

0 to

20%

of R

B V

olum

e is

typi

cal,

up to

100

%

with

exp

ansi

on v

olum

e.•

For a

naly

zed

case

s th

e R

BV

V is

app

roxi

mat

ely

equa

l to

volu

me

of

IHX

com

partm

ent a

nd P

RS

relie

f sha

ft in

NG

NP

PC

DR

des

ign

•R

adio

nucl

ide

rele

ases

from

RB

VV

dire

ctly

to e

nviro

nmen

t. •

Dep

ositi

on a

nd h

oldu

p in

rem

aini

ng R

B v

olum

e (ty

pica

lly 8

0 to

90

%) i

s ne

glec

ted

•M

odel

blo

wdo

wn,

hea

tup

and

cool

dow

n ph

ases

for c

old

brea

k D

LOFC

Slid

e64

PB

MR

TE

AM

Tech

nica

l App

roac

h (c

ontin

ued)

•In

sigh

ts fr

om p

revi

ous

HTG

R s

tudi

es in

dica

te I-

131

and

Cs-

137

are

key

to d

eter

min

ing

off-s

ite d

oses

•M

odel

cal

cula

tes

I-131

and

Cs1

37 tr

ansp

ort f

rom

fuel

rele

ase

to

offs

ite b

ound

ary

and

eval

uate

s do

se a

t the

425

m s

ite b

ound

ary

•M

odel

acc

ount

s fo

r res

uspe

nsio

n of

dep

osite

d ac

tivity

in P

HTS

du

ring

blow

-dow

n bu

t neg

lect

s pl

ate-

out a

nd re

settl

ing

of n

uclid

es

follo

win

g bl

ow-d

own

•M

odel

use

s “G

T M

HR

Pre

limin

ary

Saf

ety

Ass

essm

ent R

epor

t (D

OE

-GT-

MH

R-1

0023

0, R

ev. 0

, Sep

tem

ber 1

994)

”, Ta

ble

4.4.

3-3

to s

cale

from

cal

cula

ted

Dos

es fo

r I-1

31 a

nd C

s-13

7 to

tota

l dos

e.•

Mod

el a

ddre

sses

air

Ingr

ess

and

CO

form

atio

n du

ring

core

coo

l-do

wn

phas

e w

hen

ther

mal

exp

ansi

on o

f prim

ary

syst

em c

hang

es

to c

ontra

ctio

n•

A s

epar

ate

anal

ysis

add

ress

es a

ir in

gres

s fo

r an

assu

med

fuel

inle

tpi

pe ru

ptur

e at

top

of re

acto

r ves

sel.

Slid

e65

PB

MR

TE

AM

Rad

iolo

gica

lRel

ease

Cas

eM

atrix

•D

esig

n B

asis

Eve

nts

(DB

E) c

onsi

dere

d:–

HP

B b

reak

s in

CIP

up

to 1

00 m

m e

quiv

alen

t dia

met

er–

Loss

of f

orce

d ci

rcul

atio

n co

olin

g as

sum

ed (D

LOFC

)–

Vol

ume

of R

BV

V 1

0 %

to 1

00 %

•B

eyon

d D

esig

n B

asis

Eve

nts

(BD

BE

) con

side

red:

–H

PB

bre

aks

in C

IP fr

om 2

30 m

m to

100

0 m

m–

Loss

of f

orce

d ci

rcul

atio

n co

olin

g as

sum

ed (D

LOFC

)–

For e

valu

atio

n of

mar

gins

and

lim

iting

RB

Pre

ssur

e

•R

B D

esig

n A

ltern

ativ

es:

–C

onsi

der R

B d

esig

n al

tern

ativ

es 1

a, 1

b, 2

, 3a,

3b,

4a

and

4b

iden

tifie

d in

pre

viou

s sl

ides

Slid

e66

PB

MR

TE

AMRad

iolo

gica

l Rel

ease

Cas

e M

atrix

(c

ontin

ued)

•S

peci

al C

ases

:–

A b

ound

ing

case

(alte

rnat

ive

0) th

at s

how

s th

e im

pact

of n

o ra

dion

uclid

e re

tent

ion

by th

e R

B–

Add

ition

al A

ltern

ativ

e 4c

: A

ltern

ativ

e 4a

/b w

ith g

ross

RB

failu

re(p

uff r

elea

se o

f 100

% o

f act

ivity

in R

B a

t wor

st p

oint

in ti

me)

–10

00 m

m e

quiv

alen

t dia

met

er S

HTS

bre

ak fo

r lim

iting

RB

pr

essu

re re

spon

se (l

arge

r He

inve

ntor

y th

an P

HTS

)•

Sel

ecte

d ca

ses

cove

r a s

pect

rum

of s

cena

rios

in-w

hich

the

vario

us R

B d

esig

n fe

atur

es (b

low

-out

pan

els,

filte

rs,

pres

sure

reta

inin

g fe

atur

es) a

re s

ucce

ssfu

l and

un

succ

essf

ul in

per

form

ing

thei

r res

pect

ive

safe

ty

func

tions

suc

h as

wou

ld b

e de

velo

ped

in a

full

PR

A

Slid

e67

PB

MR

TE

AM

Maj

or A

ssum

ptio

ns

•N

orm

al o

pera

tion

core

tem

pera

ture

s an

d in

itial

radi

onuc

lide

inve

ntor

y fro

m “N

GN

P a

nd H

ydro

gen

Pro

duct

ion

Pre

conc

eptu

al

Des

ign

Stu

dy -

Con

tam

inat

ion

Con

trol (

NG

NP

-NH

S 5

0-C

C, R

ev. A

, A

pril

2008

)•

Lifto

ff an

d re

susp

ensi

on b

ased

on

mod

el c

alcu

late

d sh

ear f

orce

ra

tio•

Rad

ionu

clid

e re

leas

e fro

m fu

el d

urin

g D

LOFC

bas

ed o

n P

BM

R

exis

ting

calc

ulat

ion

for a

n ex

istin

g 50

0MW

t PB

MR

des

ign

•Th

ese

dela

yed

fuel

rele

ases

are

con

serv

ativ

e fo

r bre

ak s

izes

less

than

100

mm

due

to e

nhan

ced

conv

ectio

n co

olin

g of

the

core

whi

ch

low

ers

peak

tem

pera

ture

s du

ring

heat

-up

trans

ient

•M

ass

trans

fer b

etw

een

hot a

nd c

old

PH

TS v

olum

e an

d be

twee

n co

ld P

HTS

vol

ume

and

RB

VV

bas

ed o

n lo

ss o

f pre

ssur

e bl

owdo

wn

and

on e

xpan

sion

/con

tract

ion

durin

g he

atup

/coo

ldow

n ph

ase

Slid

e68

PB

MR

TE

AM

Maj

or A

ssum

ptio

ns (c

ontin

ued)

•N

o co

nvec

tion

or b

uoya

ncy

driv

en fl

ow a

fter d

epre

ssur

izat

ion:

–Al

l DBE

bre

aks

in P

HTS

out

side

reac

tor v

esse

l are

bre

aks

from

col

dvo

lum

e.–

Circ

ulat

or d

isch

arge

che

ck v

alve

clo

ses

at e

nd o

f blo

wdo

wn

phas

e(a

tth

e la

test

)–

Flow

pat

h fo

r buo

yanc

y dr

iven

con

vect

ion

from

cor

e to

bre

ak lo

catio

nre

quire

s flo

w re

vers

al a

nd c

ount

er-c

urre

nt fl

ow w

ith m

ultip

le u

p-do

wn

path

s–

Flow

pat

tern

s ar

e co

mpl

ex a

nd d

o no

t fav

or b

uoya

ncy

driv

en

conv

ectio

n

•B

reat

hing

rate

and

wea

ther

X/Q

for 2

4 to

96

hour

tim

e fra

me

beca

use

sign

ifica

nt c

ore

heat

up re

leas

es fr

om fu

el d

o no

t sta

rt un

til>

24 h

ours

.

Slid

e69

PB

MR

TE

AM

Ana

lytic

al M

odel

s

•Th

ree

volu

me

mod

el m

ass

flow

rate

equ

atio

ns b

ased

on

“Pre

limin

ary

Saf

ety

Info

rmat

ion

docu

men

t for

the

Sta

ndar

d M

HTG

R a

nd re

spon

ses

to N

RC

Com

men

ts”,

DO

E-H

TGR

-86-

024,

Am

endm

ent 1

3, A

ugus

t 7, 1

992.

–Is

entro

pic

chok

ed/u

n-ch

oked

equ

atio

ns u

sed

durin

g de

pres

suriz

atio

n–

Idea

l gas

ther

mal

exp

ansi

on/c

ontra

ctio

n eq

uatio

ns u

sed

afte

r de

pres

suriz

atio

n•

Dec

ay o

f I13

1 m

odel

ed in

all

volu

mes

and

in re

leas

e fro

m th

e R

BV

V.

Slid

e70

PB

MR

TE

AM

Ana

lytic

al M

odel

s (c

ontin

ued)

•Li

ft-of

f of p

late

d ou

t act

ivity

mod

el b

ased

on

“Pre

limin

ary

Saf

ety

Info

rmat

ion

Doc

umen

t for

the

Stan

dard

MH

TGR

and

Res

pons

es to

NR

CC

omm

ents

”, D

OE-

HTG

R-8

6-02

4, A

men

dmen

t 13,

Aug

ust 7

, 199

2.

–0.

2% C

s137

and

I131

lift-

off f

or a

ll sh

ear f

orce

ratio

s le

ss th

an 1

.0.

–15

% C

s137

lift-

off f

or a

ll sh

ear f

orce

ratio

s gr

eate

r tha

n 1.

0 an

d le

ss

than

30.

0–

25%

I131

lift-

off f

or a

ll sh

ear f

orce

ratio

s gr

eate

r tha

n 1.

0 an

d le

ss th

an

30.0

•D

ust r

e-su

spen

sion

mod

el b

ased

on

exis

ting

PB

MR

ana

lyse

s.–

0.02

% C

s137

resu

spen

sion

for a

ll sh

ear f

orce

ratio

s le

ss th

an 1

.0–

11%

Cs1

37 re

susp

ensi

on fo

r all

shea

r for

ce ra

tios

grea

ter t

han

1.0

and

less

than

10.

0–

34%

Cs1

37 re

susp

ensi

on fo

r all

shea

r for

ce ra

tios

grea

ter t

han

10.0

Slid

e71

PB

MR

TE

AM

Ana

lytic

al M

odel

s (c

ontin

ued)

•S

ettli

ng o

f I13

1 an

d C

s137

in th

e H

PB

was

con

serv

ativ

ely

negl

ecte

d ho

wev

er re

tent

ion

of g

as-b

orne

act

ivity

in th

e H

PB

was

m

odel

ed d

ue to

ther

mal

con

tract

ion

durin

g co

re c

ool-d

own

phas

e•

Set

tling

of I

131

and

Cs1

37 m

odel

ed in

the

RB

VV

bas

ed o

n IA

EA

TE

CD

OC

-978

,“Fu

el P

erfo

rman

ce a

nd fi

ssio

n pr

oduc

t beh

avio

r in

gas

cool

ed re

acto

rs”,

Nov

. 199

7 –

I131

dep

ositi

on c

onst

ant =

0.3

hr-1

–C

s137

dep

ositi

on c

onst

ant =

0.1

hr-1

•D

ose

calc

ulat

ion

met

hodo

logy

bas

ed o

n N

RC

Reg

. Gui

de 1

.183

, Ju

ly 2

000

•D

ose

conv

ersi

on fa

ctor

s ba

sed

on F

eder

al G

uida

nce

Rep

orts

11

(198

8) a

nd 1

2 (1

993)

Slid

e72

PB

MR

TE

AM

Ana

lytic

al M

odel

s (c

ontin

ued)

•C

onst

ant b

reat

hing

rate

of 2

.3E

-4 m

3 /s b

ased

on

the

>24

hour

br

eath

ing

rate

in N

RC

Reg

. Gui

de 1

.183

, Jul

y 20

00.

•C

onst

ant w

eath

er X

/Q o

f 2.3

E-5

s/m

3ba

sed

on 1

0% o

f the

X/Q

at

425

m in

NR

C R

egG

uide

1.4

, Jun

e 19

74.

•To

tal d

oses

from

all

radi

onuc

lides

sca

led

base

d on

“GT

MH

R

Pre

limin

ary

Saf

ety

Ass

essm

ent R

epor

t (D

OE

-GT-

MH

R-1

0023

0,R

ev. 0

, Sep

tem

ber 1

994)

”, Ta

ble

4.4.

3-3.

–To

tal t

hyro

id d

ose

from

all

radi

onuc

lides

ass

umed

to b

e a

fact

orof

2gr

eate

r tha

n I1

31 th

yroi

d do

se.

–To

tal T

ED

E d

ose

from

all

radi

onuc

lides

ass

umed

to b

e a

fact

or o

f2.5

grea

ter t

han

sum

of C

s137

and

I131

TE

DE

dos

es.

Slid

e73

PB

MR

TE

AM

Rad

iolo

gica

l Res

ults

-N

o R

B C

ase

(Alte

rnat

ive

0)

•A

ssum

es re

leas

e fro

m H

PB

is re

leas

ed to

env

ironm

ent

•Li

miti

ng b

reak

siz

e fo

r DLO

FC is

3m

m

•TE

DE

0.4

Rem

, fac

tor o

f 62.

5 le

ss th

an th

e 10

CFR

50.3

4 lim

it fo

r DB

Es;

40

% o

f PA

G li

mit

–N

o m

itiga

tion

by a

RB

is n

eede

d to

mee

t eith

er li

mit

•Th

yroi

d

10

Rem

, 2

times

PA

G li

mit

–S

ome

miti

gatio

n by

a R

B is

nee

ded

to m

eet t

he P

AG

lim

it

Slid

e74

PB

MR

TE

AM

Ana

lysi

s M

odel

-A

ltern

ativ

e 2

Rea

ctor

Cav

ity

IHX

Cav

ity

All

Oth

er C

ompa

rtm

ents

with

PH

TS H

PB S

SCs

incl

udin

g H

SS a

nd F

HSS

Pres

sure

Rel

ief

Path

way

Prim

ary

HTS

Seco

ndar

yH

TS

Rup

ture

Pan

el -

norm

ally

clo

sed

- op

ens

with

inte

rnal

pre

ssur

e -

does

not

re-c

lose

Rec

losa

ble

Dam

per

Filte

r

Slid

e75

PB

MR

TE

AM

RB

Mod

el In

put P

aram

eter

s fo

r RB

Alte

rnat

ives

•B

ase

Mod

el is

RB

Alte

rnat

ive

2•

RB

VV

= 1

00%

is 1

00,0

00 m

3•

All

RB

Alte

rnat

ives

are

sim

ulat

ed w

ith fo

llow

ing

inpu

ts:

RB

Opt

ion

Leak

Rat

e (v

%/d

)

AtP

ress

ure

(bar

-d)

RB

Vent

V

olum

eFr

actio

ns(%

of R

B)

Rec

lose

-ab

leD

ampe

rop

enin

gpr

essu

re(b

ar-a

)

Rec

lose

-ab

leD

ampe

rre

clos

ing

pres

sure

(bar

-a)

Rec

lose

-ab

leD

ampe

rFl

ow A

rea,

(m

^2)

Filte

rD

ampe

rop

enin

gpr

essu

re(b

ar-a

)

Filte

r and

Fi

lter

Dam

per

Flow

Are

a,

(m^2

)

Filte

rD

econ

ta-

min

atio

nFa

ctor

for

Cs

(%

rem

oved

)

Filte

rD

econ

ta-

min

atio

nFa

ctor

for

I (%

re

mov

ed)

1a10

00.

110

& 2

0>

0 (4

)<1

(2)

310

0 (1

)3

9995

1b10

00.

110

& 2

01.

113

incr

<1 (2

)3

100

(1)

399

952

500.

110

& 2

01.

213

incr

1.11

3 de

cr3

(3)

399

953a

500.

110

& 2

010

0 (1

)<1

(2)

31.

033

incr

399

953b

500.

150

& 1

0010

0 (1

)<1

(2)

31.

033

incr

399

954a

110

10 &

20

100

(1)

<1 (2

)3

100

(1)

399

954b

110

50 &

100

100

(1)

<1 (2

)3

100

(1)

399

95

(1)

Dam

per n

ever

ope

ns(3

)Fi

lter d

ampe

r ope

ns w

hen

recl

osab

le d

ampe

r re-

clos

es.

(2)

Dam

per n

ever

clo

ses

(4)

Dam

per a

lway

s op

en

Slid

e76

PB

MR

TE

AM

Res

ults

-RB

Alte

rnat

ives

with

RB

VV =

10%

•3

mm

bre

ak s

ize

for D

LOFC

is a

lway

s lim

iting

•M

argi

n Fa

ctor

is (P

AG

Lim

it D

ose)

/ (D

ose)

(Min

. TE

DE

or T

hyro

id)

Con

figur

atio

nLe

ak S

ize

TED

E D

ose

Thyr

oid

Dos

eM

argi

n Fa

ctor

1a 3

mm

16m

rem

400

mre

m 1

2.5

(Th)

1b 3

mm

16m

rem

400

mre

m 1

2.5

(Th)

2 3

mm

1 m

rem

20

mre

m 2

50 (T

h)3a

3 m

m1

mre

m 2

0 m

rem

250

(Th)

4a 3

mm

1.4

mre

m 3

3 m

rem

150

(Th)

Slid

e77

PB

MR

TE

AM

Res

ults

-Sen

sitiv

ity C

ases

for R

B A

ltern

ativ

es

Con

figur

atio

nR

BV

V (m

3)Le

ak S

ize

TED

E D

ose

Thyr

oid

Dos

eM

argi

n Fa

ctor

3b50

,000

3 m

m .2

mre

m4

mre

m12

504b

50,0

00 3

mm

1.3

mre

m30

mre

m16

73b

100,

000

3 m

m.0

9 m

rem

2 m

rem

2500

4b10

0,00

0 3

mm

1.1

mre

m25

mre

m20

04c

(puf

f)10

,000

3 m

m72

3m

rem

18 re

m0.

284c

(puf

f)10

0,00

0 3

mm

943

mre

m23

rem

0.22

Slid

e78

PB

MR

TE

AM

Dos

e B

ehav

ior v

s. B

reak

Siz

e fo

r Alte

rnat

ive

1a

RB

VV =

10%

0.010.

1110100

1000

1000

0

110

100

1000

Bre

ak D

iam

eter

[mm

]

Site Boundary Dose TEDE [mrem]

Dire

ctio

n of

incr

easi

ng

She

ar F

orce

s du

ring

Blo

wdo

wn

to li

ft-of

f pl

ateo

ut a

nd d

ust

Dire

ctio

n of

in

crea

sing

effe

ct o

f bl

owdo

wn

driv

ing

out t

he d

elay

ed fu

el

rele

ase

10 C

FR 5

0.34

Lim

it =

25 re

m T

ED

E

EP

A P

AG

Lim

it =

1 re

m T

ED

E

From

3m

m to

2m

m

blow

dow

n tim

e is

gr

eate

r tha

n tim

e of

fu

el re

leas

e al

low

ing

mor

e de

cay

befo

re

rele

ase

BD

BE

DB

E

Slid

e79

PB

MR

TE

AM

Tim

e D

epen

dent

Rel

ease

s of

I-13

13m

m H

PB B

reak

in C

IP –

Alte

rnat

ive

1a

RB

VV =

10%

1.0E

-04

1.0E

-03

1.0E

-02

1.0E

-01

1.0E

+00

1.0E

+01

1.0E

+02

1.0E

+03

1.0E

+04

050

100

150

Tim

e fr

om H

PB B

reak

(hrs

)

I-131 Released (Curies)

Fuel

Rel

ease

Rel

ease

from

RB

Ret

aine

d in

HPB

Ret

aine

d in

RB

VV

PH

TS F

ully

D

epre

ssur

izes

at 1

02hr

s

Slid

e80

PB

MR

TE

AM

Tim

e D

epen

dent

Rel

ease

s of

I-13

12m

m H

PB B

reak

in C

IP –

Alte

rnat

ive

1a

RB

VV =

10%

1.0E

-04

1.0E

-03

1.0E

-02

1.0E

-01

1.0E

+00

1.0E

+01

1.0E

+02

1.0E

+03

1.0E

+04

050

100

150

200

250

Tim

e fr

om H

PB B

reak

(hrs

)


Fuel

Rel

ease

Rel

ease

from

RB

Ret

aine

d in

HP

B

Ret

aine

d in

RB

VV

PH

TS F

ully

D

epre

ssur

izes

at 2

30hr

s

Slid

e81

PB

MR

TE

AM

Com

paris

on o

f I-1

31 R

elea

ses:

2m

mvs

3mm

HPB

Bre

aks

–Alte

rnat

ive

1a

01020304050

050

100

150

200

250

Tim

e fr

om H

PB B

reak

(hrs

)


RB

Rel

ease

- 2m

m

RB

Rel

ease

- 3m

m

End

of P

HTS

Dep

ress

uriz

atio

n an

d re

leas

e fro

m R

B

Slid

e82

PB

MR

TE

AM

Com

paris

on o

f Dos

e vs

. Bre

ak S

ize

for D

esig

n A

ltern

ativ

es w

ith S

ame

RB

VV =

10%

0.00

1

0.01

0

0.10

0

1.00

0

10.0

00

100.

000

1,00

0.00

0

10,0

00.0

00

110

100

1000

HPB

Bre

ak S

ize

[mm

]

Site Boundary Dose TEDE [mrem]

Alte

rnat

ive

1a/1

bAl

tern

ativ

e 2

Alte

rnat

ive

3aAl

tern

ativ

e 4a

10 C

FR 5

0.34

Lim

it =

25 re

m T

ED

E

EP

A P

AG

Lim

it =

1 re

m T

ED

E

BD

BE

DB

E

Slid

e83

PB

MR

TE

AM

Com

paris

on o

f Thy

roid

Dos

e vs

. Des

ign

Opt

ions

with

Sam

e R

BVV

= 1

0%

10100

1,00

0

10,0

00

100,

000

No

RB

Alte

rnat

ive

1a/1

bAl

tern

ativ

e 2

Alte

rnat

ive

3aAl

tern

ativ

e 2

or3a

with

failu

re o

fad

ded

feat

ures

Alte

rnat

ive

4aAl

tern

ativ

e 4a

with

late

CF

Site Boundary Thyroid Dose [mrem]

EPA

PAG

Lim

it =

5 re

m T

hyro

id

Slid

e84

PB

MR

TE

AM

Con

clus

ions

Reg

ardi

ng R

adio

logi

cal R

eten

tion

Cap

abili

ties

of th

e Ev

alua

ted

Alte

rnat

ives

•S

ome

miti

gatio

n by

a R

B is

requ

ired

to m

eet T

hyro

id P

AG

•A

ll ev

alua

ted

RB

alte

rnat

ives

pro

vide

am

ple

mar

gin

to m

eet 1

0CFR

50.3

4 an

d P

AG

s•

Alt

3a p

rovi

des

the

mos

t effe

ctiv

e ra

diol

ogic

al re

tent

ion

for D

LOFC

acr

oss

the

brea

k sp

ectru

m fo

r the

RB

VV

= 1

0% c

ases

(dos

es fo

r 3b

som

ewha

t low

er th

an

3a d

ue to

incr

ease

d vo

lum

e)•

Alt

2 pr

ovid

es c

ompa

rabl

e ra

diol

ogic

al re

tent

ion

to A

lt 3a

and

3b

for D

LOFC

for

desi

gn b

asis

bre

ak s

izes

•

Alt

1a a

nd 1

b pr

ovid

e an

indi

catio

n of

the

cons

eque

nces

of e

vent

s w

here

des

ign

feat

ures

add

ed in

Alts

2, 3

a an

d 3b

fail

(BD

BE

s)•

Alt

4a a

nd 4

b pr

ovid

e le

ss e

ffect

ive

rete

ntio

n th

an A

lt 3a

and

3b

acro

ss th

e fu

ll br

eak

spec

trum

and

less

than

Alt

2 ac

ross

the

DB

E sp

ectr

um b

ecau

seth

e R

B is

pre

ssur

ized

to d

rive

out r

adio

nucl

ides

for t

he d

urat

ion

of d

ose

calc

ulat

ion

desp

ite a

rela

tivel

y lo

w le

ak ra

te•

Alt

4c w

as d

efin

ed to

sim

ulat

e co

nseq

uenc

es o

f ass

umed

del

ayed

failu

re o

f RB

ca

pabi

lity

(BD

BE

)•

Alt

4c e

xcee

ds th

e do

se li

mits

for a

ll th

e ca

ses

anal

yzed

incl

udin

g th

eca

se w

ith n

o re

acto

r bui

ldin

g, s

ince

PH

TS is

stil

l pre

ssur

ized

at t

ime

of

RB

failu

re g

reat

er tr

ansp

ort f

or ra

dioa

ctiv

ity in

side

the

RB

and

the

HPB

Slid

e85

PB

MR

TE

AM

Cas

e 1a

Res

ults

-D

etai

ls

Ven

tM

axTi

me

to

Tim

e to

RB

Volu

me

Leak

Thyr

oid

TED

ER

BV

VD

epre

ssur

ize

Sta

rtO

ptio

nFr

actio

nSi

zeD

ose

Dos

eP

ress

ure

to 1

atm

Coo

ldow

n(m

m)

(mre

m)

(mre

m)

(bar

)(s

or h

r)(h

r)1a

0.1

1000

7.7

384.

316

.2 s

431a

0.1

230

615

1.8

63 s

431a

0.1

100

0.11

0.11

1.12

533

4 s

431a

0.1

104.

90.

291.

013

9.3

hr43

1a0.

13

398

161.

013

102

hr10

21a

0.1

232

614

1.01

323

0 hr

230

1a0.

210

006.

532

2.8

28.6

s43

1a0.

210

00.

072

0.07

81.

115

334

s43

1a0.

23

206

8.5

1.01

310

210

2

Slid

e86

PB

MR

TE

AM

Insi

ghts

–A

ltern

ativ

e 1a

Res

ults

•In

itial

inve

ntor

y fo

r rel

ease

in b

low

dow

n is

sm

all

•M

ajor

act

ivity

sou

rce

is a

ctiv

ity re

leas

e du

ring

heat

up•

Rel

ease

from

fuel

reac

hes

max

imum

at 7

2 ho

urs

for I

-131

and

100

ho

urs

for C

s-13

7•

1% o

f tot

al re

leas

e fro

m fu

el is

reac

hed

at 8

hou

rs fo

r I-1

31 a

nd 1

5 ho

urs

for C

s-13

7•

For 1

0 m

m le

ak s

ize

depr

essu

rizat

ion

is c

ompl

ete

at 9

.3 h

ours

•Fo

r gre

ater

than

10

mm

leak

siz

es th

e de

pres

suriz

atio

n is

co

mpl

ete

befo

re th

e m

ajor

rele

ase

from

the

fuel

occ

urs.

•

For g

reat

er th

an 1

0 m

m le

ak s

izes

the

dose

s ar

e lo

w b

ecau

se th

e re

leas

e fro

m fu

el is

sta

gnan

t in

the

hot v

olum

e w

ithou

t a d

rivin

gfo

rce

to th

e en

viro

nmen

t exc

ept e

xpan

sion

unt

il co

oldo

wn

star

ts.

•Fo

r gre

ater

than

10

mm

leak

siz

es th

e co

oldo

wn

star

ts a

t 43

hour

san

d is

not

per

turb

ed b

y bl

owdo

wn.

Slid

e87

PB

MR

TE

AM

Insi

ghts

-Alte

rnat

ive

1a R

esul

ts (

Con

tinue

d)

•Fo

r les

s th

an 1

0 m

m le

ak s

izes

the

blow

dow

n ex

tend

s in

to th

e he

atup

pha

se a

nd d

elay

s th

e co

oldo

wn

to 1

02 h

ours

for a

3 m

m

leak

and

230

hou

rs fo

r a 2

mm

leak

.•

For l

ess

than

10

m le

ak s

izes

the

exte

nded

blo

wdo

wn

prov

ides

a

sign

ifica

nt c

onve

ctiv

e m

echa

nism

to d

rive

radi

onuc

lides

out

of t

heH

PB

dur

ing

the

maj

or re

leas

e ph

ase.

•Th

e m

ajor

dos

es re

sult

for l

eak

size

s le

ss th

an 1

0 m

m b

ecau

se o

fth

e ex

tend

ed b

low

dow

n tra

nspo

rt of

radi

onuc

lides

to th

e R

BV

V.

•Th

e sl

ow b

low

dow

n ra

tes

allo

w m

ajor

rete

ntio

n of

radi

onuc

lides

inth

e R

BV

V.

•D

oubl

ing

the

size

of t

he R

BV

V re

duce

s th

e m

axim

um d

ose

by

alm

ost a

fact

or o

f tw

o.

Slid

e88

PB

MR

TE

AM

Des

ign

Bas

is P

eak

Pres

sure

s [1

00m

m b

reak

siz

e]

Not

es:

1. R

eclo

sabl

eda

mpe

r ope

ning

pre

ssur

e

1.11

620

3a

5.1

104a

1.12

510

3a

1.4

100

4b

1.9

504b

3.1

204a

1.07

910

03b

1.09

750

3b

1.21

3[N

ote

1]10

and

20

2

1.13

120

1b

1.13

810

1b

1.11

520

1a

1.12

510

1a

Peak

Pre

ssur

e (B

ar)

RB

VV %

Alte

rnat

ive

Slid

e89

PB

MR

TE

AM

Peak

RB

VV P

ress

ure

(con

tinue

d)

•Fo

r DB

Es

the

RB

VV

pea

k pr

essu

re is

con

trolle

d by

the

100

mm

le

ak a

nd b

y th

e R

BV

V le

ak ra

te.

•Fo

r Alts

1a

and

1b (1

00%

/d a

t 0.1

bar

d), a

nd A

lts 3

a an

d 3b

(5

0%/d

at 0

.1 b

ard)

the

RB

VV

pea

k pr

essu

re is

bel

ow 0

.125

bar

d.

•Fo

r Alt

2 th

e R

BV

V p

eak

pres

sure

is c

ontro

lled

by th

e op

enin

g pr

essu

re o

f the

recl

osea

ble

dam

per.

•Fo

r Alts

4 a

and

4 b

(1%

/d a

t 10

bar)

the

peak

pre

ssur

e ra

nges

fro

m 1

.4 b

ar to

5.1

bar

. It d

ecre

ases

with

incr

easi

ng R

BV

V v

olum

e.S

igni

fican

t tra

deof

f bet

wee

n de

sign

pre

ssur

e an

d vo

lum

e.

Slid

e90

PB

MR

TE

AMBey

ond

Des

ign

Bas

is E

vent

Mar

gins

•10

00 m

m C

old

leg

HP

B b

reak

DLO

FC is

con

trolle

d by

TE

DE

Dos

e an

d m

eets

PA

G li

mit

with

Mar

gin

Fact

or o

f 26

.•

1000

mm

Col

d le

g H

PB

bre

ak D

LOFC

sho

ws

high

R

BV

V p

ress

ure,

up

to 4

.3 b

ar fo

r Con

figur

atio

ns 1

a, 2

an

d 3a

with

10%

RB

VV

.•

With

a lo

w d

esig

n pr

essu

re to

mee

t DB

E n

eeds

the

BD

BE

s m

ight

beh

ave

mor

e lik

e a

No

RB

Con

figur

atio

n.•

Stil

l mee

ts 1

0CFR

50.3

4 Li

mits

(25

Rem

TE

DE

, 75

Rem

Th

yroi

d) b

y M

argi

n Fa

ctor

of 8

(Thy

roid

)

Slid

e91

PB

MR

TE

AM

Air

Ingr

ess

and

CO

For

mat

ion

durin

g C

old

Leg

HPB

Bre

ak D

LOFC

Coo

ldow

n Ph

ase

•A

ir In

gres

s in

to h

ot G

raph

ite re

gion

can

cau

se O

xida

tion

of

Gra

phite

at e

leva

ted

tem

pera

ture

s:–

2C+O

2 =>

2C

O +

0.1

105

kJ/K

Mol

-O2

•A

ir ca

n be

dra

wn

into

the

reac

tor v

esse

l by:

–B

uoya

ncy

driv

en fl

ow:

•H

ot H

eliu

m in

cor

e w

ants

to ri

se a

nd p

ush

cold

Hel

ium

out

thro

ugh

the

brea

k. D

raw

col

d ai

r-he

lium

mix

ture

from

he

RB

into

the

vess

el

thro

ugh

the

brea

k.

•C

lose

d C

ircul

ator

dis

char

ge c

heck

val

ve fo

rces

hot

hel

ium

and

co

ld a

ir to

flow

in c

ount

er-c

urre

nt s

tream

line

flow

pas

t eac

h ot

her

with

out m

ixin

g.

Slid

e92

PB

MR

TE

AM

Air

Ingr

ess

and

CO

For

mat

ion

durin

g D

LOFC

C

oold

own

Phas

e (c

ontin

ued)

•Fl

ow p

ath

is v

ery

com

plex

: –

For b

reak

s do

wns

tream

of c

heck

val

ve, h

eliu

m h

as to

flow

from

the

hot c

ore

up

to th

e co

re in

let p

lenu

m, t

hen

dow

n th

roug

h th

e ou

ter r

efle

ctor

to th

e co

ld lo

wer

pl

enum

, the

n up

alo

ng th

e co

re b

arre

l to

the

inle

t ple

num

at t

heto

p, th

en o

ut

thro

ugh

the

cold

inle

t pip

e to

the

brea

k be

fore

the

chec

k va

lve.

Col

d he

lium

-air

mix

ture

mus

t be

draw

n in

to th

roug

h th

e br

eak

and

flow

cou

nter

-cur

rent

to th

e ho

t hel

ium

with

out m

ixin

g –

For b

reak

s up

stre

am o

f che

ck v

alve

, hel

ium

has

to fl

ow a

gain

st b

uoya

ncy

dow

n fro

m th

e ho

t cor

e to

the

hot l

ower

cor

e pl

enum

, out

thro

ugh

the

hot o

utle

t pip

e to

the

brea

k be

twee

n th

e he

at e

xcha

nger

s or

bet

wee

n th

e se

cond

hea

tex

chan

ger a

nd th

e ci

rcul

ator

. Col

d he

lium

-air

mix

ture

mus

t be

draw

n in

thro

ugh

the

brea

k an

d flo

w c

ount

er-c

urre

nt to

the

hot c

ore

with

out m

ixin

g.–

Stre

amlin

e co

unte

r-cu

rren

t flo

w w

ithou

t mix

ing

in th

ese

com

plex

geo

met

riear

eve

ry u

nlik

ely.

Buo

yanc

y fo

rce

in h

ot h

eliu

m is

1/7

th o

f sam

e te

mpe

ratu

rebu

oyan

cy in

air

beca

use

of lo

w h

e de

nsity

.

Slid

e93

PB

MR

TE

AM

Air

Ingr

ess

and

CO

For

mat

ion

durin

g Sm

all C

old

Bre

ak

DLO

FC C

oold

own

Phas

e (c

ontin

ued)

–C

ontra

ctio

n du

ring

cool

dow

n ph

ase

•C

oold

own

phas

e do

es n

ot s

tart

until

43

hour

s, lo

nger

for

brea

ks <

10

mm

.•

Con

tract

ion

of a

tmos

pher

ic p

ress

ure

Hel

ium

in P

HTS

dra

ws

cold

Air-

Hel

ium

mix

ture

into

col

d P

HTS

vol

ume

thro

ugh

brea

k an

d m

ixes

with

Hel

ium

in c

old

PH

TS v

olum

e.•

Con

tract

ion

of h

ot P

HTS

vol

ume

draw

s co

ld v

olum

e A

ir-H

eliu

m m

ixtu

re in

to h

ot v

olum

e.•

Oxy

gen

reac

hing

hot

cor

e ca

n re

act w

ith h

ot g

raph

ite.

•C

alcu

latio

ns w

ith 3

vol

ume

mod

el s

how

that

less

than

2 M

ol

of C

O is

pro

duce

d by

300

hou

rs d

ue to

con

tract

ion

effe

cts

alon

e.

Slid

e94

PB

MR

TE

AM

Air

Ingr

ess

thro

ugh

Fuel

Inle

t Pip

e B

reak

at

Top

of R

eact

or V

esse

l

•C

onfig

urat

ion

and

Sol

utio

n–

One

Fue

l Tub

e pe

r 120

ºse

ctor

of c

ore

(3 tu

bes)

–65

mm

dia

met

er w

ith ri

bs to

gui

de 6

0 m

m d

iam

eter

peb

bles

–Po

stul

ate

Fuel

Tub

e Br

eak

at T

op o

f Ves

sel

–Bl

owdo

wn

of c

old

PHTS

vol

ume

thro

ugh

core

will

tend

to c

ool c

ore

dow

n. P

HTS

is d

epre

ssur

ized

at 1

000

seco

nds.

–H

ot H

eliu

m in

cor

e w

ill ris

e th

roug

h Fu

el T

ube

and

draw

col

d Ai

r-H

eliu

mm

ixtu

re in

from

RB

.–

Cou

nter

-cur

rent

flow

in s

traig

ht v

ertic

al tu

be–

Forc

e B

alan

ce:

Buo

yanc

y fo

rce

= Fr

ictio

n lo

sses

–Vo

lum

e Fl

ow b

alan

ce:

Volu

met

ric fl

ow ra

te o

f Hel

ium

out

= A

ir in

–So

lved

with

dou

ble

itera

tion

on fr

ictio

n fa

ctor

s fo

r Hel

ium

and

for A

ir

Slid

e95

PB

MR

TE

AM

Air

Ingr

ess

thro

ugh

Fuel

Tub

e B

reak

at T

op o

f Ve

ssel

(con

tinue

d)•

Ass

umpt

ions

–At

end

of B

low

dow

n as

sum

e H

eliu

m te

mpe

ratu

re in

cor

e an

d Ai

r-H

eliu

m m

ixtu

re in

RB

are

both

at 1

000

ºC–

Ass

ume

pure

air

in R

B (c

onse

rvat

ive,

max

imiz

es b

uoya

ncy)

–A

ssum

e 8

ribs

of 2

mm

hei

ght i

n fu

el tu

be–

Ass

ume

all a

ir en

terin

g co

re re

acts

with

Gra

phite

to fo

rm C

O

•R

esul

ts–

Hel

ium

upf

low

is la

min

ar a

nd o

ccup

ies

86%

of f

uel t

ube

flow

are

a–

Air d

ownf

low

is tu

rbul

ent a

nd o

ccup

ies

14%

of f

uel t

ube

flow

are

a–

Volu

met

ric u

pflo

w =

dow

nflo

w =

0.0

0151

m3/

s–

Air

mas

s flo

w =

air

ingr

ess

rate

= 4

.18E

-04

kg/s

–R

epla

ces

Hel

ium

in c

ore

and

uppe

r and

low

er p

lenu

m in

10.

3 ho

urs

–C

O fo

rmat

ion

rate

= 0

.028

3 M

ol C

O/s

–R

eact

ion

ener

gy =

1.56

mW

(Com

pare

: Dec

ay H

eat =

5 M

W a

t 3 h

rs)

Slid

e96

PB

MR

TE

AM

Air

Ingr

ess

thro

ugh

Fuel

Tub

e B

reak

at T

op o

f Ve

ssel

(con

tinue

d)

•C

oncl

usio

ns–

Air

Ingr

ess

thro

ugh

rupt

ured

fuel

tube

is s

mal

l–

CO

reac

tion

ener

gy is

neg

ligib

le c

ompa

red

to d

ecay

he

at–

Ana

lysi

s is

con

serv

ativ

e (p

ure

air i

n R

B, e

qual

te

mpe

ratu

res,

all

O2

reac

ts to

form

CO

)

Slid

e97

PB

MR

TE

AM

Insi

ghts

Reg

ardi

ng P

ress

ure

Cap

aciti

es

•R

B s

houl

d w

ithst

and

the

high

er o

f the

blo

w-d

own

pres

sure

of 1

.15

bar o

r the

ope

ning

pre

ssur

e of

th

e re

-clo

seab

le d

ampe

r or f

ilter

dam

per f

or

DB

Es.

•R

B s

houl

d w

ithst

and

abou

t 5 b

ar if

requ

ired

to b

e in

tact

for B

DB

Es.

Slid

e98

PB

MR

TE

AM

Lim

itatio

ns a

nd A

reas

for F

urth

er R

esea

rch

•C

oncl

usio

ns re

gard

ing

radi

olog

ical

rete

ntio

n ca

pabi

lity

of e

valu

ated

RB

optio

ns a

re s

ubje

ct to

lim

itatio

ns d

ue to

–La

ck o

f des

ign

deta

ils a

nd a

ssoc

iate

d fu

ll sc

ope

PR

A m

odel

–N

eed

to e

valu

ate

diffe

rent

HP

B b

reak

loca

tions

and

a fu

ller s

et o

f lic

ensi

ng

basi

s ev

ents

–N

eed

to c

onsi

der t

he im

pact

of n

atur

al c

onve

ctio

n on

cor

e te

mpe

ratu

res

durin

g sm

all l

eaks

(2-1

0mm

)–

Lack

of a

fully

inte

grat

ed m

echa

nist

ic s

ourc

e te

rm m

odel

–N

eed

to c

onsi

der t

he q

uant

itativ

e fa

ilure

pro

babi

litie

s of

var

ious

des

ign

feat

ures

as

wel

l as

RB

stru

ctur

al c

apab

ility

to w

ithst

and

load

s fro

m a

full

set o

f lic

ensi

ng

basi

s ev

ents

–La

ck o

f a fu

ll un

certa

inty

ana

lysi

s in

the

sour

ce te

rm a

nd c

onse

quen

cem

odel

ing

•Th

ese

limita

tions

sho

uld

be a

ddre

ssed

in th

e C

once

ptua

l and

Pre

limin

ary

Des

ign

Stag

es o

f the

NG

NP

•A

deta

iled

mec

hani

stic

cod

e w

ith in

tegr

ated

PH

TS a

nd R

B m

odel

s, in

plac

e of

the

sim

plifi

ed 3

vol

ume

mod

el, w

ill be

requ

ired

for t

hene

xt p

hase

of

the

anal

ysis

.

Slid

e99

PB

MR

TE

AM

Inte

grat

ed E

valu

atio

n of

Alte

rnat

ive

Rea

ctor

Bui

ldin

g D

esig

ns

Slid

e10

0P

BM

R T

EA

M

App

roac

h to

Eva

luat

ion

of A

ltern

ativ

es

•E

very

alte

rnat

ive

is p

resu

med

to b

e ca

pabl

e of

mee

ting

all n

on-

safe

ty fu

nctio

nal r

equi

rem

ents

(mos

tly g

eom

etry

and

stre

ngth

)•

App

roxi

mat

e pr

essu

re tr

ansi

ents

are

cal

cula

ted

for s

ever

al b

reak

size

s an

d lo

catio

ns•

Eng

inee

red

feat

ures

(rup

ture

pan

els,

filte

rs, e

tc) a

nd p

ossi

ble

chan

ges

in b

uild

ing

geom

etry

and

stre

ngth

requ

irem

ents

are

es

timat

ed fo

r eac

h al

tern

ativ

e•

Rad

ionu

clid

e re

tent

ion

by in

here

nt a

nd e

ngin

eere

d fe

atur

es is

es

timat

ed fo

r eac

h al

tern

ativ

e, a

nd s

ite b

ound

ary

dose

s ar

e es

timat

ed•

Cap

ital c

osts

are

est

imat

ed fo

r eac

h al

tern

ativ

e

Slid

e10

1P

BM

R T

EA

M

Eval

uatio

n fo

r Nor

mal

Ope

ratio

n C

riter

ia

•N

orm

al O

pera

tion

Goa

ls–

Goo

d op

erat

or a

cces

s du

ring

oper

atio

n an

d m

aint

enan

ce–

Min

imum

ope

rato

r exp

osur

e to

radi

atio

n an

d ot

her h

azar

ds–

Min

imum

nor

mal

ope

ratin

g re

leas

e of

radi

onuc

lides

to o

ffsite

pub

lic•

Eva

luat

ion

of A

ltern

ativ

es–

1a:

Ade

quat

e ac

cess

and

con

trol o

f exp

osur

e; M

ay n

ot p

rovi

de c

ontro

l of a

ir ac

tivat

ion

prod

ucts

, and

may

requ

ire la

rge

HV

AC

flow

-> s

core

: 9

–1b

: A

dequ

ate

acce

ss a

nd c

ontro

l of e

xpos

ure;

ven

t pat

h no

rmal

ly is

olat

ed

from

env

ironm

ent i

mpr

oves

con

trol o

f air

activ

atio

n pr

oduc

ts a

nddo

es n

ot

requ

ire la

rge

HV

AC

flow

-> s

core

: 10

–2:

Sam

e as

1b

-> s

core

: 10

–3a

: R

educ

ed le

ak ra

te re

quire

s ai

rlock

s an

d ot

her i

mpe

dim

ents

to o

pera

tor

acce

ss, o

ther

wis

e, s

imila

r to

2->

sco

re:

8–

3b:

Sam

e as

3a

-> s

core

: 8–

4a a

nd 4

b: p

ress

ure

rete

ntio

n re

quire

s ad

ditio

nal f

eatu

res

that

impe

de o

pera

tor

acce

ss->

sco

re: 3

Slid

e10

2P

BM

R T

EA

M

Eval

uatio

n fo

r Inv

estm

ent P

rote

ctio

n C

riter

ia

•In

vest

men

t Pro

tect

ion

Goa

ls–

Low

forc

ed o

utag

e ra

te a

nd d

urat

ions

–Lo

w ri

sk o

f eve

nts

that

can

cau

se d

amag

e to

pla

nt s

yste

ms,

st

ruct

ures

, and

com

pone

nts

(SSC

s)–

Low

risk

of e

vent

s th

at c

an re

sult

in s

igni

fican

t pla

nt o

utag

e tim

e–

Acce

ptab

ly lo

w ri

sk o

f pla

nt w

rite-

off

•E

valu

atio

n of

Alte

rnat

ives

–Th

ose

with

mor

e sy

stem

s an

d co

mpo

nent

s ha

ve a

n in

crea

se in

forc

edou

tage

s an

d an

incr

ease

in d

ownt

ime

–Th

ose

with

mor

e st

ringe

nt le

akag

e re

quire

men

ts h

ave

an in

crea

se in

outa

ge d

urat

ions

–Le

ak ti

ght o

ptio

ns ta

ke m

ore

time

to re

cove

r fro

m ra

dion

uclid

e re

leas

es

Slid

e10

3P

BM

R T

EA

M

Eval

uatio

n fo

r Saf

ety

Eval

uatio

n fo

r Saf

ety

Crit

eria

:H

PB L

eaks

/Bre

aks

Res

pons

e to

Pre

ssur

e Tr

ansi

ent

•H

PB

Lea

ks a

nd B

reak

s D

esig

n G

oal

–M

aint

ain

stru

ctur

al g

eom

etry

of r

eact

or a

nd R

CC

S d

urin

g pr

essu

retra

nsie

nt•

Eva

luat

ion

of A

ltern

ativ

es–

1a:

Ven

ted

desi

gn re

sults

in s

urvi

vabl

e pr

essu

re tr

ansi

ent w

ithhi

ghre

liabi

lity-

> s

core

: 10

–

1b:

Rup

ture

pan

els

incr

ease

pre

ssur

e tra

nsie

nt s

light

ly, h

elp

to is

olat

e an

d pr

otec

t HVA

C->

sco

re:

9–

All o

ther

alte

rnat

ives

sam

e sc

ore

as 1

b

Slid

e10

4P

BM

R T

EA

M

Eval

uatio

n fo

r Saf

ety

Crit

eria

: H

PB L

eaks

/Bre

aks

Rel

ativ

e to

Rad

ionu

clid

e R

eten

tion

•H

PB

Lea

ks a

nd B

reak

s D

esig

n G

oal

–P

rovi

de re

tent

ion

of ra

dion

uclid

es re

leas

ed fr

om H

PB

and

fuel

•E

valu

atio

n of

Alte

rnat

ives

–S

core

s co

rrela

te to

radi

onuc

lide

rele

ases

and

offs

ite d

oses

ove

rsp

ectru

m o

f LBE

s–

3a a

nd 3

b re

ceiv

e hi

ghes

t sco

res

base

d on

the

resu

lts o

f the

ra

diol

ogic

al re

leas

e st

udy,

alth

ough

they

will

have

seq

uenc

es in

whi

chth

e fil

ters

do

not o

pera

te w

ith th

e hi

gher

dos

es o

f 2, 1

a, a

nd 1

b–

4a a

nd 4

b ha

ve e

vent

s w

ith re

leas

e of

bot

h pr

ompt

and

del

ayed

fuel

sour

ce te

rm d

riven

by

pent

-up

non-

cond

ensa

ble

heliu

m

Slid

e10

5P

BM

R T

EA

M

Eval

uatio

n fo

r Eva

luat

ion

for S

afet

y C

riter

ia:

Seis

mic

•S

eism

ic D

esig

n G

oal

–M

aint

ain

geom

etry

dur

ing

desi

gn b

asis

sei

smic

eve

nts

and

capa

bilit

yfo

r sei

smic

BD

BE

s•

Eva

luat

ion

of A

ltern

ativ

es–

All a

ltern

ativ

es d

esig

ned

for s

eism

ic w

ith m

argi

n–

All

mus

t res

pond

with

cap

abilit

y fo

r bey

ond

SS

E e

vent

s –

Thos

e w

ith fe

wer

filte

rs a

nd d

ampe

r gra

ded

high

er–

Thos

e w

ith h

ighe

r lea

k tig

htne

ss m

ore

susc

eptib

le to

sei

smic

-indu

ced

crac

ks a

nd le

akag

e

This

crit

eria

impa

cted

by

degr

ee o

f em

bedm

ent

Slid

e10

6P

BM

R T

EA

M

Eval

uatio

n fo

r Eva

luat

ion

for S

afet

y C

riter

ia:

Hyd

roge

n / P

roce

ss H

azar

ds•

Hyd

roge

n P

roce

ss H

azar

d D

esig

n G

oals

–P

rote

ct R

B in

tern

als

from

sho

ck/b

last

wav

es

–P

rote

ct R

B in

tern

als

from

che

mic

al c

loud

s/to

xici

ty•

Eva

luat

ion

of A

ltern

ativ

es–

1a:

Ext

erna

l pre

ssur

e lo

adin

g du

e to

hyd

roge

n ex

plos

ion

coul

d be

grea

ter t

han

pres

sure

tran

sien

t loa

ding

; O

pen

vent

pat

h m

ay w

eake

nre

sist

ance

to h

ydro

gen

expl

osio

n->

Sco

re:

5–

1b:

Sam

e as

1a

-> S

core

: 5

–2:

Gre

ater

than

ope

n al

tern

ativ

es 1

a an

d 1b

-> S

core

: 8–

3a a

nd 3

b: M

ore

robu

st b

uild

ing

mea

ns h

azar

ds n

ot li

kely

to c

ontro

l or

impa

ct d

esig

n->

Sco

re:

9–

4a a

nd 4

b: E

ven

mor

e ro

bust

bui

ldin

g, w

ithou

t ven

t or f

ilter

s, o

ffers

grea

test

resi

stan

ce to

hyd

roge

n ev

ent h

azar

ds->

Sco

re:

10

This

crit

eria

impa

cted

by

degr

ee o

f em

bedm

ent

Slid

e10

7P

BM

R T

EA

M

Eval

uatio

n fo

r Sec

urity

/ A

irpla

ne C

rash

C

riter

ia•

Sec

urity

Goa

ls–

Prev

ent m

alev

olen

t int

erve

ntio

n fro

m im

pact

ing

plan

t saf

ety

or

oper

atio

n–

Pro

vide

pro

tect

ion

of R

B in

tern

als

to a

irpla

ne c

rash

•E

valu

atio

n of

Alte

rnat

ives

–1a

and

1b

have

ope

n ve

nt p

athw

ay ->

sco

re:

5–

All

othe

r alte

rnat

ives

-> S

core

: 10

This

crit

eria

may

be

impa

cted

by

degr

ee o

f em

bedm

ent

Slid

e10

8P

BM

R T

EA

M

Eval

uatio

n fo

r Cos

t Crit

eria

•C

ost G

oals

–M

inim

ize

plan

t con

stru

ctio

n co

st–

Min

imiz

e op

erat

ing

cost

s•

Eva

luat

ion

of A

ltern

ativ

es–

1a:

Min

imal

feat

ures

and

load

s ->

Sco

re: 1

0–

1b:

Mod

est a

dditi

onal

feat

ures

-> S

core

: 9

–2:

Add

ition

al d

ampe

r and

filte

r, no

t sig

nific

ant c

ost d

river

s ->

Sco

re: 8

–3a

: S

igni

fican

t cos

t pen

alty

for r

educ

ed le

ak ra

te.

Stru

ctur

alde

sign

pro

babl

y no

t con

trolle

d by

pre

ssur

e lo

ads

-> S

core

: 5

–3b

: S

imila

r to

3a, b

ut w

ith a

dditi

onal

cos

t for

exp

ansi

on v

olum

e->

Sco

re:

4–

4a:

Larg

e co

st in

crem

ent f

or lo

w le

ak ra

te, p

ress

ure

reta

inin

g ca

pabi

lity

->S

core

: 2–

4b:

Sim

ilar t

o 4a

, but

with

mod

est r

educ

tion

in c

ost d

ue to

low

er p

ress

ure

load

; ad

ditio

nal c

ost f

or e

xpan

sion

vol

ume

-> S

core

: 1

This

crit

eria

may

be

impa

cted

by

degr

ee o

f em

bedm

ent

Slid

e10

9P

BM

R T

EA

M

Eval

uatio

n fo

r Lic

ensi

ng C

riter

ia

•Li

cens

ing

Goa

ls–

Mee

t sta

tuto

ry li

mits

for L

BEs

at s

ite b

ound

ary

–M

eet P

AG fo

r LBE

sfo

r no

evac

uatio

n at

site

bou

ndar

y•

Crit

eria

is n

ot to

tally

und

er o

ur c

ontro

l and

judg

men

t at t

his

time

is

high

ly s

ubje

ctiv

e•

Link

s to

the

mar

gins

of t

he s

afet

y cr

iteria

and

our

abi

lity

to c

redi

bly

pres

ent a

nd d

efen

d ou

r non

-LW

R s

afet

y de

sign

app

roac

h•

Eva

luat

ion

of A

ltern

ativ

es–

1a:

In p

art b

ecau

se o

f eve

nts

with

prio

r gra

phite

mod

erat

ed re

acto

rs,

air i

ngre

ss w

ill be

an

issu

e w

ith th

is o

pen

desi

gn–

All

othe

r ven

ted

optio

ns a

re ra

nked

bas

ed o

n th

eir d

ose

mar

gins

–O

ptio

ns 4

a an

d 4b

hav

e th

e po

tent

ial f

or th

e hi

ghes

t con

sequ

ence

sequ

ence

s

Slid

e11

0P

BM

R T

EA

M

Eval

uate

d Sc

ore

of A

ltern

ativ

es

4959 7 10 510101b

98109 8 9 88102

105109 10 8 9693a

104109 10 8 9683b

62109 5 6 10434a

410510 7 10 5991a4b

Alte

rnat

ive

6Li

cens

abilit

y

1C

apita

l and

ope

ratin

g co

st

10Se

curit

y / a

ircra

ft cr

ash

9 5 6 10

Safe

ty re

quire

men

tsH

BP le

aks/

brea

ks p

ress

ure

resp

onse

HBP

leak

s/br

eaks

radi

onuc

lide

rele

ase

Seis

mic

Hyd

roge

n/Pr

oces

s H

azar

ds

4In

vest

men

t pro

tect

ion

requ

irem

ents

3N

orm

al o

pera

ting

requ

irem

ents

Slid

e11

1P

BM

R T

EA

M

Wei

ghte

d Sc

ore

of A

ltern

ativ

es

9090

150

150

135

6060

15Li

cens

abilit

y

100

251012 8 5 10510

Wei

ght

749

225

50108

56 50 5050100

1b

872

200

100

108

64 45 80401002

813

125

100

108

80 40 9030903a

778

100

100

108

80 40 9030803b

568

50100

108

40 30 100

20304a

771

250

50120

56 50 5045901a4b

Alte

rnat

ive

543

Tota

l

25C

apita

l and

ope

ratin

g co

st

100

Secu

rity

/ airc

raft

cras

h

108

40 30 100

Safe

ty re

quire

men

tsH

BP le

aks/

brea

ks p

ress

ure

resp

onse

HBP

leak

s/br

eaks

radi

onuc

lide

rele

ase

Seis

mic

Hyd

roge

n/Pr

oces

s H

azar

ds

20In

vest

men

t pro

tect

ion

requ

irem

ents

30N

orm

al o

pera

ting

requ

irem

ents

Slid

e11

2P

BM

R T

EA

M

Eval

uatio

n Su

mm

ary

of V

ente

d vs

Pres

sure

-R

etai

ning

RB

Opt

ions

•In

gen

eral

the

vent

ed o

ptio

ns a

re s

uper

ior t

o th

e pr

essu

re-r

etai

ning

opt

ions

bas

ed o

n th

e fo

llow

ing

cons

ider

atio

ns;

–G

reat

er c

ompa

tibilit

y w

ith a

non

-con

dens

able

and

iner

t prim

ary

cool

ant

–V

entin

g of

the

prim

ary

cool

ant i

nven

tory

to a

tmos

pher

e w

ith o

r w

ithou

t filt

ratio

n el

imin

ates

a d

rivin

g fo

rce

for s

ubse

quen

t ra

dion

uclid

e tra

nspo

rt of

the

dela

yed

fuel

rele

ase

–B

ette

r or c

ompa

rabl

e ra

dion

uclid

e re

tent

ion

capa

bilit

y fo

r al

tern

ativ

es w

ith ra

nge

of d

esig

n fe

atur

es–

Low

er c

osts

for m

itiga

tion

prov

ided

–E

asie

r and

less

cos

tly to

eng

inee

r int

erfa

ces

with

RC

CS

, S

HTS

, FH

SS

, HS

S, a

nd o

ther

NH

SS

and

aux

iliary

sys

tem

s

Slid

e11

3P

BM

R T

EA

M

Eval

uatio

n Su

mm

ary

of V

ente

d R

B O

ptio

ns

•H

ighe

st ra

ting

of in

tegr

ated

eva

luat

ion

for a

ltern

ativ

es e

xam

ined

was

Alt

2 fo

llow

ed b

y A

lt 3a

•B

oth

alte

rnat

ives

pro

vide

sup

erio

r rad

ionu

clid

e re

tent

ion

capa

bilit

yfo

r des

ign

basi

s H

PB

bre

aks

with

DLO

FC th

an p

ress

ure

reta

inin

g al

tern

ativ

es; A

lt 3b

clo

sely

follo

wed

by

3a is

sup

erio

r in

term

sof

radi

onuc

lide

rete

ntio

n to

all

eval

uate

d al

tern

ativ

es a

cros

s th

e en

tire

HP

B b

reak

spe

ctru

m in

clud

ing

AO

Os,

DB

Es,

and

BD

BE

s•

Alts

2, 3

a, a

nd 3

b ar

e ex

pect

ed to

hav

e gr

eate

r lic

ensa

bilit

yth

anei

ther

of t

he o

pen

vent

ed o

ptio

ns 1

a an

d 1b

•

Ano

ther

alte

rnat

ive

for f

utur

e st

udy

is a

pas

sive

recl

osab

leda

mpe

rw

ithou

t a fi

lter (

Alt

1c).

This

is e

xpec

ted

to h

ave

dela

yed

fuel

rele

ase

rete

ntio

n ca

pabi

litie

s ap

proa

chin

g th

at o

f Alt

2

Slid

e11

4P

BM

R T

EA

M

Rad

ionu

clid

e R

eten

tion

Allo

catio

ns fo

r R

ecom

men

ded

RB

Alte

rnat

ive

•R

esul

ts o

f the

radi

onuc

lide

rete

ntio

n st

udy

show

that

all

the

eval

uate

d al

tern

ativ

es p

rovi

de s

uffic

ient

mar

gins

to o

ffsite

dos

elim

its b

ased

on

pass

ive

radi

onuc

lide

rete

ntio

n w

ithin

the

fuel

and

HP

B•

Add

ed e

ngin

eere

d fe

atur

es s

uch

as fi

lters

and

re-c

losa

ble

dam

pers

ad

d ad

ditio

nal m

argi

ns•

This

stu

dy c

onfir

ms

that

radi

olog

ical

rete

ntio

n is

not

a re

quire

dsa

fety

func

tion

but r

athe

r a s

uppo

rtive

saf

ety

func

tion

for t

he N

GN

Pre

acto

r bui

ldin

g•

It is

reco

mm

ende

d th

at a

RB

des

ign

goal

be

set f

or ra

dion

uclid

e re

tent

ion

capa

bilit

y of

a fa

ctor

of 1

0 re

duct

ion

in re

leas

es fr

om th

e H

PB

for I

-131

and

Cs-

137

for D

BE

HP

B b

reak

s

Slid

e11

5P

BM

R T

EA

M

Ope

n Is

sues

and

Add

ition

al E

ngin

eerin

g St

udie

s•

Con

clus

ions

rega

rdin

g ra

diol

ogic

al re

tent

ion

capa

bilit

y of

eva

luat

ed R

B op

tions

ar

e su

bjec

t to

limita

tions

due

to–

Lack

of d

esig

n de

tails

and

ass

ocia

ted

full

scop

e P

RA

mod

el–

Nee

d to

eva

luat

e di

ffere

nt H

PB

bre

ak lo

catio

ns a

nd a

fulle

r set

of l

icen

sing

bas

is

even

ts–

Nee

d to

con

side

r the

impa

ct o

f nat

ural

con

vect

ion

on c

ore

tem

pera

ture

s du

ring

smal

l le

aks

(2-1

0mm

)–

Lack

of a

fully

inte

grat

ed m

echa

nist

ic s

ourc

e te

rm m

odel

–N

eed

to c

onsi

der t

he fa

ilure

pro

babi

litie

s of

var

ious

des

ign

feat

ures

as

wel

l as

RB

st

ruct

ural

cap

abili

ty to

with

stan

d lo

ads

from

a fu

ll se

t of l

icen

sing

bas

is e

vent

s–

Lack

of a

full

unce

rtain

ty a

naly

sis

in th

e so

urce

term

and

con

sequ

ence

mod

elin

g•

Thes

e lim

itatio

ns s

houl

d be

add

ress

ed in

the

Con

cept

ual a

nd P

relim

inar

yD

esig

n St

ages

of t

he N

GN

P su

ppor

ted

by th

e N

GN

P PR

A•

Key

cha

lleng

es id

entif

ied

for r

eact

or b

uild

ing

desi

gn–

Opt

imiz

atio

n of

RV

ven

ted

volu

me

dim

ensi

ons

vspe

rform

ance

and

cos

t–

Unk

now

ns re

gard

ing

need

ed p

rote

ctio

n ag

ains

t hyd

roge

n pr

oces

s ha

zard

s–

Unk

now

ns re

gard

ing

need

ed p

rote

ctio

n ag

ains

t phy

sica

l sec

urity

thre

ats

–S

yste

ms

inte

ract

ions

issu

es a

ssoc

iate

d w

ith S

HTS

pip

ing

pene

tratio

n R

B w

alls

•Pe

ndin

g a

mor

e th

orou

gh d

esig

n ite

ratio

n, D

DN

sw

ill be

form

ulat

ed le

adin

g to

po

tent

ial t

echn

olog

y de

velo

pmen

t prim

arily

in th

e fu

el, r

eact

or,a

nd H

PB

Slid

e11

6P

BM

R T

EA

M

Futu

re C

onsi

dera

tion

of B

uild

ing

Inte

ract

ions

•K

ey re

quire

men

t for

the

RB

is to

pro

vide

phy

sica

l sep

arat

ion

of th

eN

HS

S fr

om e

vent

s an

d ha

zard

s as

soci

ated

with

the

HP

S, P

CS

, an

d B

OP

faci

litie

s•

SH

TS p

ipin

g pr

ovid

es s

truct

ural

link

age

betw

een

RB

and

adj

acen

t bu

ildin

gs•

Nee

d to

inve

stig

ate

“sys

tem

s in

tera

ctio

ns”i

nvol

ving

–Fa

ults

in H

PS o

r PC

S pr

opag

atin

g in

to R

B–

Ext

erna

l eve

nts

for w

hich

RB

is p

rote

cted

but

oth

er b

uild

ings

are

not

ca

usin

g ad

vers

e in

tera

ctio

ns•

Nee

d to

pro

vide

hig

h co

nfid

ence

of n

o ad

vers

e in

tera

ctio

ns fo

r de

sign

bas

is e

vent

s•

May

lead

to id

entif

icat

ion

mul

tiple

larg

e H

PB

bre

aks

for B

DB

Es

durin

g th

e C

once

ptua

l Des

ign

PR

A•

Key

cha

lleng

e in

the

next

pha

se o

f the

des

ign

•Is

sue

may

com

plic

ate

the

appr

oach

to e

mbe

dmen

t

Documents

NGNP and Hydrogen Production Conceptual Design Study