An Integrated Risk-Informed Evaluation Model for Reactor Core, Fuel Performance, and Systems Analysis

CASL Industry Council Meeting, April 4-5, 2017, Charleston, SC

Hongbin ZhangLight Water Reactor Sustainability (LWRS)

Risk-Informed Safety Margin Characterization (RISMC)CASL AMA

Idaho National Laboratory

RISMC Industry ApplicationsTopics for Discussion

DOE LWRS Background RISMC Industry Applications Overview LB-LOCA Demonstration Analysis LOTUS Integrated Evaluation Model Approach Plant Demonstration – 4 Loop PWR

1. Core Design2. Fuel Performance3. Systems Analysis4. UQ and Risk Assessment5. Results

Fuels Performance under LOCA– Modeling Needs– Future Work

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Background

The existing NPP fleet is facing economic challenges due to:– Low natural gas price– Rapid deployment of renewable energy sources

Additionally, the existing NPP fleet is facing regulatory challenges, e.g.– 10 CFR 50.46c– 10 CFR 50.69

Reducing fuel cost is one area that contributes to the improved economic viability and enhanced safety of the existing fleet– Higher burnup of the fuel– Optimized fuel and loading pattern design– Accident tolerant fuel– Load following and flexible operating strategies

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Multi-physics, multi-scale analysis of core/fuel/systems is desired

Nuclear Industry Stakeholders

Methods

Risk

Reliability

Tools

Grizzly –Materials, Aging

RAVEN –Advanced Risk

Assessment

RELAP-7 –Next Generation Systems Code

External Hazards –Seismic, Flood,

Wind

Data

Assessment, Verification &

Validation

Experiments

RISMCRisk-Informed Safety Margin Characterization

RISMC Industry Applications –Assisting Margin Management & Sustainability through Realistic Demonstrations

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3-Step Demonstrations:– Problem Definition– Preliminary Demo– Full Analysis Demo

Multi-physics Multi-scale

Nuclear Analysis

Dynamic / Combined Analysis

IndustryApplications

Uncertainty Quantification

1. ECCS/LOCA (50.46c)Core and Fuel

Multi-physics Systems Demo

2. External Hazards –Combined Events Demo

Seismic & Flooding

Helping Demonstrate Game Changers in Delivering the Nuclear Promise

50.69AccidentTolerant

Fuel

Risk-Informed Thinking

RISMC Industry Applications

Purpose:

Risk-Informed analysis of realistic, relevant industry problems, with accurate representation of margins for the long term benefit of nuclear assets.

Strategy:

Develop industry application demonstrations in collaboration with the nuclear industry;

Align demonstrations with RISMC methods and tools capabilities;

Follow existing industry application structure, starting with existing (legacy codes) which will be replaced with advanced tools as they become available.

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A Demonstration of Decision Making Tools and Methods

RISMC Margin Quantification and Risk Assessment Paradigm

Risk Simulation

Scenario Generator

Failure Modes

Operational Rules

Physics Simulation(full physics coupling)

Core Design Automation

Fuel/Clad Performance

System Analysis

Core Design Optimization

Decision Making

Margin and Uncertainty Quantification

Scenarios

Parameters

PCTECR CWO

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Industry Application LOCA –Integrated Cladding/ECCS Performance

Motivation– Based upon recent experiments, NRC proposed new regulations (10 CFR 50.46c)

• Peak-clad temp. and embrittlement oxidation more restrictive than current limits• LWRS program will help industry by using RISMC to demonstrate safety margins

for loss-of-coolant-accident (LOCA) analysis including emergency core cooling system (ECCS) performance under realistic plant conditions

– Coupled analysis has core physics, cladding behavior, thermal-hydraulics, and scenario-based risk analysis in order to quantify safety margin

Potential Impact: – Cost of re-analysis

• ~65 License Amendment Requests, > $100M (U.S. plants)• 7-year implementation plan

– Loss of margin– Increased fuel costs– Operation flexibility impact– Increased complexity

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Integrated Cladding/ECCS PerformanceLB LOCA Analysis Demonstration

Proposition: – Re-analysis can be used to better understand/manage margins– Efficient assessment of margins (through advanced methods/tools)– Opportunity for reload design and operations processes improvements

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Proposed 50.46c Limits

A targeted value proposition for vendor and/or owner-operator stakeholders Goal: Industry adoption via pilot/demos/tools

Assessment Optimization

5. Core Design Optimization

(CD-O)

RISMC Application for LOCA Analysis LOTUS (LOCA Toolkit U.S.)

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1. Core Design Automation

(CD-A)2.

Fuels/Clad Performance

(FP)

3. Systems Analysis

(SA)

4. Risk Assessment

(RA) Methods

Assembly/pin power, exposure, etc.

Assembly/pin power, exposure, etc.

RELAP Output

Optimized Fuel/Core Design

Probabilistic PCT/ECR margins,Data analysis

Fuel/Core Design

BEPU, Reduced Order Models, Surrogate Models, Limit Surfaces, etc.

Fuel/Core Design

Industry Application

Demonstration(LOTUS) Pin power, Fuels performance data,

safety limits

Stored energy, Rod internal pressure, Clad oxidation and perforation, etc.

What is LOTUS?

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LOTUS is an automation tool to integrate reactor core, fuel and systems analysis– Increase efficiency and reduce analysis cost– Reduce chance of error in data stream management

New approach to propagate uncertainties consistently in a multi-physics, multi-scale environment– Uncertainties are propagated directly from uncertain

design and model parameters– Interactions between various parameters are directly

resolved LOTUS will utilize existing simulation tools and

advanced tools being developed Risk-informed decision making enabling toolkit

5. Core Design Optimization

(CD-O)

LOTUS Toolkit – Baseline and Advanced

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1. Core Design Automation

(CD-A)2.

Fuels/Clad Performance

(FP)

3. Systems Analysis

(SA)

4. Risk Assessment

(RA) Methods

BISON (A)

PHISICS (B)

RELAP5 (B)

RAVEN (A) (B)

VERA-CS (A)

Industry Application

Demonstration(LOTUS) RELAP-7 (A)

FRAPCON/FRAPTRAN (B)

(B) – Baseline(A) – Advanced

To Be Developed

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1. Core Design Automation

13HE-LL Design HE-LL-O Design

PWR Core Design –Generic Design Based on STP

PWR Core Design –Generic Design Based on STP

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Coupled RELAP5/PHISICS– 1 TH channel per assembly– Boundary conditions at

lower and upper plena Developed PWR core similar to

STP core– 3.8 GWth– 14 feet Westinghouse core

Design Criteria– 18 month cycle– “HE-LL” design

• High energy/low leakage– Equilibrium assumed

after 8 cycles– Enrichment 4.2%-4.6 %– Fresh/1/2/burned map from a

similar PWR– IFBA distribution obtained by

optimization process

Demo PWR Cross section generation

1/8 Core in HELIOS 33 libraries generated

– 29 fuel assemblies– blanket– 1 radial reflector– 1 top, 1 bottom reflector

Homogenized to 8 energy groups 4 tabulation dimensions

– Fuel temperature (3)– Moderator density (4)– Boron concentration (3)– Burn-up (4)

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Explicit IFBA coating

Fuel and Assembly Design

17 x 17 pins 25 GT positions Top and bottom blanket

– (assumed low enriched)

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Assembly example with 128 IFBA Bottom of active fuel

Top of active fuel

7’’ top blanket 2.6% enriched

7’’ bottom blanket 2.6% enriched

4.2%/4.6% enriched fuel

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2. Fuel/Clad Performance

Fuel Design Data –Based on STP-4 Loop PWR

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Fuel rod characteristics STP FSAR Ref.Rod diameter 0.374 in (0.94996 cm) Table 4.1-1Rod diametral gap (2xgap size) 0.0065 in (0.01651 cm) Table 4.1-1Cladd thikness 0.0225 in (0.05715 cm) Table 4.1-1Pellet diameter 0.3225 in (0.81915 cm) Table 4.1-1Cladding material ZIRLOTM / Optimized ZIRLOTM Chap 4.1-1Helium pressure level BOL 1000 psia (68.948 bar) Chap 4.2-21

IFBA coating Assumed AssumedIFBA loading 1.57mg/inch B10

Instrumnent tube wall thickness 0.02 in (0.508 mm) Chap 4.1-1Guide tube wall thickness 0.0185 in (0.04699 cm) Table 4.1-1Guide tube inner diameter 0.442 in (1.12268 cm) Fig 4.2-7BGuide tube outer diameter 0.482 in (1.22428 cm) Fig 4.2-7B

Control rod diameter 0.366 in (0.92964 cm) Table 4.3-1Control rod cladding thickness 0.0185 in (0.04699 cm) Table 4.3-1Control rod active length 158.9 in (403.606 cm) Amendment page 5-6

Fuels/Clad Performance (Baseline)

Fuel mechanics– RELAP5-3D includes rupture model and ballooning model – But we need detailed analysis of fuel rods’ behaviors such as the fission

gas released, rod internal pressure, and fuel-cladding mechanical interaction, cladding H content etc., FRAPCON

The power history data is automatically retrieved by LOTUS from the core design results going into a FRAPCON input

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Power history for the hot rod (one assembly)

Cladding hydrogen content vs. rod average burn-up (all assemblies)

fresh 1-burned 2-burned

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3. Systems Analysis

Safety Analysis – Core Hydraulic Homogenization and Heat Structures

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An existing RELAP5 PWR model is modified to analyze typical four-loop PWR cores:– A core hydraulic homogenization

is performed– Heat structures for the hot

assembly in each group are connected to the hot channel in that group (2 sets of heat structures for each assembly)

RELAP5 nodalisation for a typical four-loop PWR

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An existing RELAP5 PWR model is modified to analyze the typical four-loop PWR cores:– Reactor Vessel

• Downcomer• Bypass• Lower/Upper plena• Core• Upper head

– Reactor coolant system• 4 primary loops• Secondary side up to turbine

governor valves– ECCS

• Low pressure injection (LPI) • High pressure injection (HPI)

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4. Uncertainty Quantification and Risk Assessment

The industry standard Code Scaling, Applicability and Uncertainty (CSAU) methodology was used in our LB-LOCA analyses.

A CSAU based Best Estimate Plus Uncertainty (BEPU) LB-LOCA analysis methodology was developed.

– Select a Plant Model, Accident Scenario and Safety Metrics– Identify Relevant Physical Phenomena and Uncertain Parameters with Their

Probability Distribution Functions– Develop Simulation Models for LB-LOCA Analysis– Random Sampling of Uncertain Parameters and Performing RELAP5-3D Runs– Determining the Limiting Values of the Safety Metrics

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Uncertainty Quantification Methodology

• In current licensing practice Wilk’s nonparametric statistical methods are used, where only a small number of runs (59, 124, etc. ) is needed

• In LOTUS we use a Monte Carlo approach to better quantify margins and to take advantage of high performance computing

Risk Assessment – Baseline BEPU

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LB-LOCA with a double-ended guillotine break in a cold leg BEPU analysis

– PIRT Reduces set of parameters with high importance Automatically mapped parameters from fuel performance and core design

– cladding pre-transient hydrogen up-take contents, rod internal pressure, gap gas mole fraction, power distribution, etc.

7 LOCA start times in cycle and maneuver 1000 Monte Carlo samples for each start time

Distribution of Parameter Uncertainties

Parameter

PDF type

Min

Max

Comments

Reactor thermal power

Normal

0.98

1.02

Multiplier

Reactor decay heat power multiplier

Normal

0.94

1.06

Multiplier

Accumulator pressure

Normal

-0.9

1.1

Multiplier

Accumulator liquid volume (m3)

Uniform

-0.23

0.23

Additive

Accumulator temperature (K)

Uniform

-11.1

16.7

Additive

Subcooled multiplier for critical flow

Uniform

0.8

1.2

Multiplier

Two-phase multiplier for critical flow

Uniform

0.8

1.2

Multiplier

Superheated vapor multiplier for critical flow

Uniform

0.8

1.2

Multiplier

Fuel thermal conductivity

Normal

0.93

1.07

Multiplier

Average core coolant temperature (K)

Normal

-3.3

3.3

Additive

Film boiling heat transfer coefficient

Uniform

0.7

1.3

Multiplier

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5. LOTUS Coupling - Results

Risk Assessment of Two Core Designs: (1) HE-LL vs. (2) HE-LL-O results

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(1)

(2)

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ECR Results by Fuel Type (Burnup) for a HE-LL Core Reload and Design Strategy (Cont’d)

An Advanced Toolkit to Inform Design and Operations

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Understand Simulation Envelope

Optimization

Better Economics /Operational Flexibility

Similar concept can be used for other safety/operational limitsto support ATF and BU extension

Important fuel performance issues under LOCA– Corrosion– Hydrides– Fuel fragmentation and pulverization at high BU– Axial relocation – crumbled fuel relocates in the ballooned

region– Fuel rod ballooning and burst– Fuel dispersal:

• Fuel rod burst will determine the amount of fuel dispersal • Size of the fuel fragments and rupture opening, axial

mobility of fuel fragments need to be quantified. • Fuel coolant interaction• Hydraulic and mechanic effects• Radiological release 30

Issues with Fuel Performance under LOCA

BEPU methods need to be developed for coupled fuels performance simulations to quantify the issues shown in the previous slide

Models Available but NOT in the Fuels Performance Codes (FRAPTRAN/BISON)

– Axial relocation

Models to be developed– Rupture opening size– Fuel pulverization size versus burnup– Model for extent of fuel dispersal – Consequences of fuel dispersal (Fuel coolant interaction,

Hydraulic and mechanic effects, Radiological release)31

Modeling Needs

The loosely coupling approach is used in the steady state fuel performance analysis– VERA-CS coupling with FRAPCON– VERA-CS coupling with BISON

The power histories, coolant flow conditions are obtained by VERA-CS and input files for FRAPCON and BISON prepared by LOTUS

LOTUS will carry out the uncertainties propagation in the VERA-CS/FRAPCON and VERA-CS/BISON calculations

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Steady State Fuel Performance and Systems Analysis

The strongly coupling approach is required in the transient fuel performance analysis– RELAP5-3D/FRAPTRAN– RELAP5-3D/BISON– RELAP-7/BISON

The coupled codes are run concurrently to provide BEPU analyses

LOTUS will carry out the uncertainties propagation in the RELAP5/FRAPTRAN, RELAP5/BISON, RELAP-7/BISON calculations.

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Transient Coupled Fuel Performance and Systems Analysis

The LOTUS framework is being developed to provide a multi-physics simulation capability to better quantify safety margins. The LOTUS capability has been demonstrated by integrating baseline tools such as PHISICS, FRAPCON and RELAP5-3D.

Future work will focus on coupled systems/fuels performance simulations under steady state and transient conditions by incorporating advanced simulations tools such as VERA-CS, BISON, and eventually RELAP-7 when it becomes available.

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Summary and Future LOTUS Development Work

Helping to Sustain National Assets

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An Integrated Risk-Informed Evaluation Model for Reactor Core, Fuel Performance, and Systems AnalysisRISMC Industry Applications�Topics for DiscussionBackgroundRISMC Industry Applications –�Assisting Margin Management & Sustainability through Realistic DemonstrationsRISMC Industry ApplicationsSlide Number 6Industry Application LOCA –�Integrated Cladding/ECCS PerformanceIntegrated Cladding/ECCS Performance�LB LOCA Analysis DemonstrationRISMC Application for LOCA Analysis LOTUS (LOCA Toolkit U.S.)What is LOTUS?LOTUS Toolkit – Baseline and AdvancedSlide Number 12PWR Core Design – �Generic Design Based on STP PWR Core Design – �Generic Design Based on STP Demo PWR Cross section generationFuel and Assembly DesignSlide Number 17Fuel Design Data – �Based on STP-4 Loop PWRFuels/Clad Performance (Baseline) Slide Number 20Safety Analysis – Core Hydraulic Homogenization and Heat StructuresRELAP5 nodalisation for a typical four-loop PWRSlide Number 23Uncertainty Quantification MethodologyRisk Assessment – Baseline BEPUSlide Number 26Risk Assessment of Two Core Designs: (1) HE-LL vs. (2) HE-LL-O results ECR Results by Fuel Type (Burnup) for a HE-LL Core Reload and Design Strategy (Cont’d)An Advanced Toolkit to Inform Design and OperationsIssues with Fuel Performance under LOCAModeling NeedsSteady State Fuel Performance and Systems AnalysisTransient Coupled Fuel Performance and Systems AnalysisSummary and Future LOTUS Development WorkSlide Number 35