Advanced Reactors: Code and Research Needs

Bill CorwinOffice of Advanced Reactor Technologies

Office of Nuclear EnergyU.S. Department of Energy

NRC Standards Forum

September 8, 2016Bethesda, Maryland

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Gen IV Non-Light-Water-Cooled Advanced Reactors Are Being Developed to Augment LWRs

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Sodium Fast Reactor

Integral part of the closed fuel cycle• Can either burn actinides or breed

fissile materialDetailed design of ASTRID (France),

JSFR (Japan), PGSFR (Korea) are proceeding, BN-1200 is under development (Russia)

Planned start-up of BN-800 (Russia) R&D focus

• Analyses and experiments to demonstrate safety approaches

• High burn-up minor actinide bearing fuels

• Develop advanced components and energy conversion systems

550°C

SFR is one of the systems the US is actively pursuing.

AFR 100AFR-100 Advanced Sodium-cooled Fast Reactor

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Goal - reach outlet temperature of 1000°C, with near-term focus on 750-850°C

Reference configurations are the prismatic and the pebble bed• Designed to be “walk away safe”

High temperature enables non-electric applications

R&D focus on materials and fuelsHTTR in operation in Japan HTR-PM demonstration plant

under construction in China

Very High Temperature Reactor

750-1000°C

VHTR is one of the systems the US is actively pursuing.

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Lead Fast Reactor

Lead is not chemically reactive with air or water and has lower coolant void reactivity

Three design thrusts:• European Lead Cooled Fast

Reactor (Large, central station)• Russian BREST-OD-300 (Medium

size)• SSTAR (Small Transportable

Reactor)R&D focus on materials corrosion

and safety 480-800°C

Candidate technology for both reactor and accelerator driven system.

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Molten Salt Reactor

High temperature system• High temperature enables non-

electric applicationsOn-line waste managementDesign Options

• Solid fuel with molten salt coolant• Fuel dissolved in molten salt

coolantKey technical focus

• Neutronics• Materials and components• Safety and safety systems• Liquid salt chemistry and

properties• Salt processing

Conceptual MSR Design

700-800°C

Recent interest by several industrial concerns

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Gas-cooled Fast Reactor

High temperature, inert coolant and fast neutrons for a closed fuel cycle• Fast spectrum enables extension of

uranium resources and waste minimization

• High temperature enables non-electric applications

• Non-reactive coolant minimizes material corrosion

Very advanced system• Requires advanced materials and fuels

Key technical focus:• SiC clad carbide fuel • High temperature components and

materials

850°C

The big question is how to remove decay heat passively.

GFR Reference Design

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Supercritical Water-Cooled Reactor

510-625°C

Merges GEN-III+ reactor technology with advanced supercritical water technology used in coal plants

Operates above the thermodynamic critical point (374°C, 22.1 MPa) of water

Fast and thermal spectrum optionsBeing examined for RPV and pressure tube

designsKey technology focus:

• Materials, water chemistry, radiolysis• Thermal hydraulics and safety to address

gaps in SCWR heat transfer and critical flow databases

• Fuel qualificationContinued fundamental R&D before launching a prototype and in-pile testing.

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Integrated Light-Water-Cooled SMRs May Be Deployed Sooner and Address Licensing Issues

NuScale 45 MWeNuScalePower

mPower 125 MWeBabcock & Wilcox

IRIS 300 MWeWestinghouse

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ASME Codes & Standards Are Integral Parts of Nuclear Plant Construction and Licensing

ASME Codes & Standards• Provide rules for safe construction & in-service inspection

Owner/Operator• Applies for plant license• Responsible for Code Implementation• Provides input on needed Code rules

Department of Energy• Develops technical input for Code updates &

improvementsNuclear Regulatory Commission

• Reviews and may endorse Code rules to facilitate licensing

ASME Codes & Standards Provide Valuable Pathway to Apply Results from Advanced Reactor R&D

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ASME Nuclear Code Coverage Is Largely in Sections III and XI

Division 1 - Metallic Components, Low & Hi Temp

Divisions 2 & 3 – Concrete Containments and Containments for Spent Fuel and High Level Waste

Division 5 – Construction and Design Rules for High Temperature Reactors (issued November 2011)

Rules for Inservice Inspection of Nuclear Power Plant Components

(non-nuclear)

Rules for Construction of Power Boilers

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ASME Section III Treats Metallic Materials for Low & High Temperatures Separately

Allowable stresses for LWR & low-temperature advanced reactor components not time dependent• < 700°F (371°C) for ferritic steel and

< 800°F (427°C) for austenitic matls

At higher temps, materials behave inelastically and their allowable stresses are explicit functions of time & temperature • Must consider time-dependent

phenomena such as creep, creep-fatigue, relaxation, etc.

THTR Steam Generator

PWR RPV

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•Sec III Div 5 contains construction and design rules for high-temperature reactors, including gas-, metal- & salt-cooled reactors

•Covers low temperature metallic components, largely by reference of other portions of Sec III

•Covers high-temperature metallic components explicitly, including former

•Sec III, Subsections NG (Core Supports) & NH (Elevated Temperature Components)•Relevant Code Cases addressing time-dependent behavior

•Also includes rules for graphite & ceramic composites for core supports & internals for first time in any international design code

•Numerous technical issues in Div 5 have been identified. Some have been and others are being addressed

Sec III Div 5, Specifically Developed to Address High Temperature Reactors,

Was Issued in November 2011

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Materials and High Temperature Design Methods in Div 5 Need Updating*

Weldments• Weldment evaluation methods, metallurgical & mechanical

discontinuities, transition joints, tube sheets, validated design methodology

Aging & environmental issues• Materials aging, irradiation & corrosion damage, short-time over-

temperature/load effects Creep and fatigue

• Creep-fatigue (C-F), negligible creep, racheting, thermal striping, buckling, elastic follow-up, constitutive models, simplified & overly conservative analysis methods

Multi-axial loading• Multi-axial stresses, load combinations, plastic strain

concentrations

*Based on Multiple DOE, NRC & National Lab Reviews of High Temperature Reactor Issues over Past 40 Years

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Materials allowables• Elevated temperature data base & acceptance criteria, min vs ave

props, effects of melt & fab processes, 60-year allowables Failure criteria

• Flaw assessment and LBB procedures Analysis methods and criteria

• Strain & deformation limits, fracture toughness, seismic response, core support, simplified fatigue methods, inelastic piping design, thermal stratification design procedures

NRC Endorsement of Div 5 & associated Code Cases• Alloy 617 • Strain Limits for Elevated Temp Service Using E-PP Analysis• Creep-Fatigue at Elevated Temp Using E-PP Analysis

DOE Advanced Reactor Technology R&D Supports Resolution of These Issues Plus Development &

Qualification of Data Required for Design

Div 5 Materials and High Temperature Design Methods Need Updating (cont)

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High Priority ASME Code Committee Actions Endorsed by BNCS and DOE

Topics 2017Edition

Beyond 2017

New simplified analysis methods (EPP) that replace current methods based onlinear analysis

X

Adequacy of the definition of S values used for the design of Class B components, which is based on extrapolated properties at 100,000 hours, in light of application to 500,000 hours design

X

Construction rules for “compact” heat exchanges XIncorporation of new materials such as Alloy 617 and Alloy 709 (austenitic

stainless)A617 A709

Pursuit of “all temperature code” XComplete the extension of Alloy 800H for 500,000 hr-design XComplete the extension of SS304, 316 for 500,000 hr-design XComplete the extension of Grade 91 for 500,000 hr-design XThermal striping XDevelop design by analysis rules for Class B components (including compact

HX)X

Component classification (Refer back to ANS 53 classification rules), including assessment of: Is Class B really necessary?

X

Add non-irradiated and irradiated graphite material properties X

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Technical Bases for Code Rule Development of Graphite and Ceramic

Composites Continuing to Expand

Graphite used for core supports in HTGRs, VHTRs and FHRs• Maintain core geometry and protect fuel• Includes current and future nuclear graphites

Special graphite considerations for Code rules• Lack of ductility• Need for statistically set load limits• Requires irradiation and oxidation data

Ceramic composites (e.g. SiC-SiC) for internals & controls for gas, liquid-metal & salt cooled systems• Very high temperature and irradiation resistance• Dosemax > 100 dpa, Tmax ≥ 1200°C• Materials specification, design, properties, testing,

examination, and reporting rules developing

Core Tie Rod

Graphite Core

Supports

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ASME Section XI Covers ISI for LWRs and Advanced Non-LWRs

Division 1 addresses LWRsDivision 2 addresses gas-

cooled reactorsDivision 3 addresses liquid-

metal-cooled reactorsDivisions 2 and 3 are

published but out of date and are being updated• Div 2 will use the Reliability

and Integrity Management (RIM) approach

• Div 3 is providing alternate rules incorporating System Based Code (SBC)

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Section XI Division 2 Is Being Rewritten Using Risk Informed Methodology

Reliability and Integrity Management (RIM):

Those aspects of the plant design process that are applied to provide an appropriate level of reliability of SSCs and a continuing assurance over the life of the plant that such reliability is maintained.

Includes design features important to reliability performance such as design margins, selection of materials, testing and monitoring, maintenance, repair and replacement, pressure and leak testing, and In-service Inspection (ISI).

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Section XI Division 2 Will Comprise RIM-Based Platform-Independent Body with Coolant-Specific Appendices

Div 2 was originally developed to address HTGR system needs, but has now been redirected and expanded

Will cover water-, gas-, metal-, and salt-cooled systems Includes:

• Scope of SSCs coverage and their damage mechanisms

• Establishing plant- and SSC-level reliability targets and RIM Strategies to achieve them

• Evaluation of uncertainties in reliability performance• Monitoring SSC reliability performance and updating

RIM program20

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Margin accumulated but how much is not clear.

Target reliability is determined first.

LOAD

DESIGN

INSPECTION

TOTALINTEGRITY

MATERIAL

etc…...

APPROPRIATE

EXCESSIVE

LOAD

DESIGN

INSPECTION

TOTALINTEGRITY

MATERIAL

etc…...

APPROPRIATE

Present System Based Code

Design to required reliabilityDesign to required reliability

Margin exchangeMargin exchange

Expansion of technical optionsExpansion of technical options

Update of reliability evaluation

Update of reliability evaluation

Alternate SBC Rules for Division 3Will Allow Exchange and Optimization of Design and Inspection Margins

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ASME Code Case for Alternate Liquid Metal Reactor ISI Rules for Division 3 Developed Jointly with JSME

SBC process consists of Stage Iand Stage II evaluations.

• A structural reliability evaluationconsiders component level structural integrity and probability of failure of the component under design basis conditions. The contribution of in-service inspections is not taken into account

• A safety-related evaluation of the ability to detect a flaw that ensures that the plant can be safely shut down before the flaw reaches the maximum acceptable size. If a flaw is not detectable, additional margins in structural integrity will be required as penalty.

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Target reliability is determined first.

LOAD

DESIGN

INSPECTION

TOTALINTEGRITY

MATERIAL

etc…...

APPROPRIATE

System Based Code

Design to required reliabilityDesign to required reliability

Margin exchangeMargin exchange

Update of reliability evaluation

Update of reliability evaluation

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Sec III Div 5 Construction Rules for HTRs were issued in 2011 • Covers high temperature metals and design methods, as well as

graphite and ceramic composites

ASME and DOE jointly established priority list for needed Div 5 rules improvements in 2015 and are pursuing them

NRC has begun to assess ASME Div 5 for endorsement • Very important since predecessor ASME docs never endorsed• Will facilitate HTR design process and enhance regulatory surety

Alternate, risk-informed approached for inspection and overall system based Code are being developed for advanced reactors within Sec XI Div 2 & 3

DOE materials program providing technical basis to address ASME Code improvements and NRC endorsement concerns

ASME Nuclear Code Activities Are Critical for Advanced Reactor Development

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QUESTIONS?

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Reviews for Advanced Reactors Found Shortcomings in High-Temp Metals & High-Temp Design Methodology (HTDM)

NRC/ACRS Review of Clinch River Breeder Reactor in mid-1980’s [1]

GE’s PSID for PRISM 1986 – NRC Generated PSER in 1994 [2] ORNL Review for NRC of ASME Code Case N-47 (now NH and

Div 5A) in 1992 [3] NRC Review and Assessment of Codes and Procedures for

HTGR Components in 2003 [4] DOE-funded ASME/LLC Regulatory Safety Issues in

Structural Design Criteria Review of ASME III NH in 2007 [5] NRC-sponsored Review of Regulatory Safety Issues for

Elevated Temperature Structural Integrity for Next Generation Plants in 2008 [6]

These reviews cumulatively identified over 40 individual concerns, but can be summarized under 8 key areas

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References for High Temperature Reactor Materials and Design Methods Reviews

1. Griffen, D.S., “Elevated-Temperature Structural Design Evaluation Issues in LMFBR Licensing,” Nuclear Engineering and Design, 90, (1985), pp. 299-306

2. NUREG-1368 “Preapplication Safety Evaluation Report for the Power Reactor Innovative Small Module (PRISM) Liquid-Metal Reactor,” Feb. 1994

3. NUREG/CR-5955, Huddleston, R.L. and Swindeman, R.W., “Materials and Design Bases Issues in ASME Code Case N-47,” ORNL/TM-12266, April 1993

4. NUREG/CR-6816, Shah, V.N., S. Majumdar, and K. Natesan,“Review and Assessment of Codes and Procedures for HTGR Components,” ATL-02-36, June 2003.

5. O’Donnell, W. J., and D. S. Griffin, “Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR and Gen-IV,” ASME-LLC STP-NU-010, Dec. 2007

6. O’Donnell, W.J., Hull, A.B., and Malik, S., “Historical Context of Elevated Temperature Structural Integrity for Next Generation Plants: Regulatory Safety Issues in Structural Design Criteria of ASME Section III Subsection NH,” Proceedings of 2008 ASME Pressure Vessel and Piping Conf., PVP2008-61870, July 2008

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Alloy 617 Code Case currently in review process• Advanced gas reactor heat exchangers & steam

generators up to 950°C and 100,000 hrs• Low-temperature Code Case (T < 427°C) submitted

May 2014 and high-temperature in Sept 2015• Anticipate inclusion in 2019 edition of Sec III Div 5

Alloy 709 selected for Code qualification• Will provide improved

performance, design envelop, and cost reduction for LMRs

• Roughly double existing creep strength of existing stainless steels in Sec III Div 5

• Detailed qualification plan prepared and testing begun

Additional High-Temperature Alloys, Now Being Qualified, Will Provide Additional

Options for Nuclear Construction

Creep-fatigue crack 617

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Improved design rules, based on elastic-perfectly plastic analysis, proposed for strain limits & creep-fatigue• Critical for very high temperatures where no

distinction exists between creep and plasticity– Current rules invalid at very high

temperatures– Will enable simplified methods for Alloy 617

> 1200°F (649°C)• E-PP analysis addresses ratchetting & shakedown• Avoids stress classification

Yield strength is a “pseudo” strength given by the limiting design parameter, e.g. stress for 1% inelastic strain

The Rapid Cycle (RC) is limiting case that bounds the real Steady Cyclic (SC) solution

Improved Components of High-Temperature Design

Methodology Being Developed

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Two-bar tests can simulate combined thermal transients and sustained pressure loads that can generate a ratchet (progressive deformation) mechanism during creep-fatigue, relaxation, elastic follow-up, etc.• Validation of the E-PP model under varying effects of thermal path and

mean stress

Advances in High Temperature Design Methodology Are Being Validated

through Key Features Tests

Ti

To

∆T

p

rF=prPressurized cylinder with

radial thermal gradient

To Ti

F=prEquivalent Two-

bar model

Equal deformations

Pressure stress in vessel wall represented by total load on bars;

Through-wall temperature gradient represented by temperature difference between bars

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ASME LLC Tasks Funded by DOE-NE

Task 1 Verification of Allowable Stresses in ASME Section III, Subsection NH with Emphasis on Alloy 800H and Modified 9Cr-1Mo Steel

Task 2 Regulatory Safety Issues in Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR and GEN IV

Task 3 Improvement of Subsection NH Rules for Modified 9Cr-1Mo Steel Task 4 Updating of ASME Nuclear Code Case N-201 to Accommodate

the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development

Task 5 Collect Available Creep - Fatigue Data and Study Existing Creep - Fatigue Evaluation Procedures

Task 6 Review of Current Operating Conditions Allowable Stresses in ASME Section III Subsection NH

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ASME LLC Tasks Funded by DOE-NE (cont)

Task 7 Evaluate ASME Code Considerations for High Temperature Reactor Intermediate Heat Exchangers

Task 8 Creep and Creep - Fatigue Crack Growth at Structural Discontinuities and Welds

Task 9 Update Section III Division 1 Subsection NH – Simplified Elastic and Inelastic Methods

Task 10 Update Section III Division 1 Subsection NH – Alternative Simplified Creep – Fatigue Design Methods

Task 11 New Materials for Section III Division 1 Subsection NH Task 12 NDE and ISI Technology for High Temperature Gas-

Cooled Reactors (NRC-Funded) Task 13 Recommend Allowable Stress Values for 800H Task 14 Correct Allowable Stress Values for 304 and 316 Stainless

Steel