CASL: Consortium for the Advanced Simulation of Light Water Reactors

A DOE Energy Innovation Hub

Achievements in Addressing Challenges Facing the Light Water Reactor Industry

Dave Pointer, PhD. Deputy Lead, Thermal Hydraulics Methods

Oak Ridge National Laboratory

for

Dave Kropaczek, PhD.CASL Chief Scientist

North Carolina State University

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CASL was the first DOE Energy Innovation Hub• Established by Former DOE Energy

Secretary Steven Chu• Modeled after the scientific management

characteristics of Manhattan Project and AT&T Bell Labs:– Addressing critical problems– Combines basic and applied research with engineering– Integrated team to take discovery to application

• Four Hubs are in operation

For more info: http://energy.gov/science-innovation/innovation/hubs

“Multi-disciplinary, highly collaborative teams ideally working under one roof to

solve priority technology challenges”

– Steven Chu

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CASL’s Mission is to Provide Leading-Edge M&S Capabilities to Improve the Performance of Operating LWRs

VISIONPredict, with confidence, the performance and assured safety of nuclear reactors, through comprehensive, science-based M&S technology deployed and applied broadly by the U.S. nuclear energy industry

GOALS• Develop and effectively apply modern virtual reactor technology• Provide more understanding of safety margins while addressing

operational and design challenges• Engage the nuclear energy community through M&S• Deploy new partnership and collaboration paradigms

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CASL is a National Laboratory, Industry, University Partnership

Core Physics, Inc.

CASL Founding Partners

CASL Contributing Partners

International Collaborators

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CASL Scope: Develop and apply a “Virtual Reactor” to assess fuel design, operation, and safety criteria• Deliver improved predictive simulation

of Light Water Reactors– Focus on fuels, vessel, internals– First five year focus on PWRs, broadened to BWR and Light

Water Small Modular Reactors• Execute work in five technical

focus areas to:– Equip the Virtual Reactor with necessary physical models and

multi-physics integrators– Build the Virtual Reactor with a comprehensive, usable, and

extensible software system – Validate and assess the Virtual Reactor models with self-

consistent quantified uncertainties

Focus on Addressing Challenge Problems to Drive Development and Demonstration

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Virtual Environment for Reactor Applications

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M&S Key Aspect - Multi-Physics Coupling

With rigorous representation of physics feedback, simulations yield higher confidence predictions of core performance

Thermal Hydraulics

Neutronics

Fuel Performance

Fluid Temperature

Fluid Density / Void BISON

MPACT

COBRA-TF

Neutronic Power

Gamma Heating

Boron Concentration

Fuel Temperature

Clad Heat Flux

Clad Surface Temperature

MAMBA

Crud Thickness

Crud Composition (Boron)

Crud Thermal Resistance

Chemistry

APR1400 VERA Simulation

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CASL Challenge Problems are Focused on Key Industry Reactor Performance Areas

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Application – Watts Bar 2 Initial Reactor Startup Predictions• Completed and independently verified the VERA

model in early 2016, based on input from TVA and Westinghouse

• Performed zero power physics tests calculations in March, three months prior to startup– Critical boron concentration– Control bank reactivity worths– Isothermal Temperature Coefficient

• Comparisons also made with results from Westinghouse design methods

• Performed full core SDM calculations in support of requests from TVA

• Provided increased confidence in predictions from NRC licensed design codes

HZP BOC Fission Rate Distribution in WB2

Worked performed by: J. Ritchie1 A. Godfrey2

1 Tennesee Valley Authority 2 Oak Ridge National Laboratory

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– Dec. 2015 – Fuel Load– May 23, 2016 – Initial criticality– June 3, 2016 – On the power grid; Begin power ascension testing– August 30, 2016 – Reactor trip from 99% power (transformer fire)– September 30, 2016 – Power Ascension Testing completed– October 19, 2016 – Full power commercial operation

Watts Bar Nuclear Plant – Unit 2

• Spring City, TN• First new nuclear plant in

U.S. since 1996 (WBN1)• Traditional four-loop

Westinghouse PWR• 3411 MWth initial rated

thermal power• Current burnup:

~50 EFPD

Image courtesy of TVA

Notable Dates:

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Power History for PAT

Turbine generator coupling making excessive noise (5/28/2016)

Automatic trip and safety injection on steam pressure low (6/5//2016)

Automatic trip from Lo-Lo level in number 4 steam generator(6/20/2016)

Planned 10% load rejection

Manual trip due to low steam generator levels caused by a loss of feedwater flow from main feedwater pump (8/23/2016)

Planned loss of offsite power trip from 30% (7/14/2016)

Loss of bushing cooling due to excessive hydrogen leak, unable to exceed 75% power

Turbine trip from a main bank transformer failure (8/30/16)

50% load rejection

Turbine trip (6/26/2016)

Planned trip from outside of MCR (8/3/2016)

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0.6% 0.8%

3.1%

-0.7%

3.1%

0.8% 1.0%

-0.7%

0.9%

-10%

-8%

-6%

-4%

-2%

0%

2%

4%

6%

8%

10%

D C B A SD SC SB SA Total

Bank

Wor

th D

iffer

ence

(%)

RCCA Bank

Startup Results*Measured MPACT

DifferenceShift

DifferenceInitial Critical Boron Concentration (ppmB) 1089 -14 -2

Isothermal Temperature Coefficient (pcm/ºF) -5.31 -0.15 --

Total Worth Error < 1%

*Measurements courtesy of TVA

Control Bank Worths

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Zero Power Criticality Measurements • Criticality Measurements taken at hot-zero-power

conditions following shutdowns• Includes various Bank D positions and transient Xenon-

135 conditions

-14 -16 -17 -17-20 -20

-22 -23-18

-13

-2-6 -7 -6

-9 -9-12 -13

-8-3

-50

-40

-30

-20

-10

0

10

20

30

40

50

Boro

n Co

ncen

tratio

n Di

ffere

nce (

ppm

)

MPACT

SHIFT

= -18 ± 3.4 ppm

= -8 ± 3.6 ppm

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VERA Boron Concentrations

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

650

700

750

800

850

900

950

1,000

1,050

1,100

1,150

5/23 5/30 6/6 6/13 6/20 6/27 7/4 7/11 7/18 7/25 8/1 8/8 8/15 8/22 8/29 9/5 9/12 9/19 9/26

Core

Pow

er (%

)

Solu

ble B

oron

Con

cent

ratio

ns (p

pmB)

Date

MeasuredVERAPower

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VERA Runtime Performance• Each quarter-core calculation has used 4234 cores on

OLCF’s Eos supercomputer

• Analysis :– 31 jobs– 3,047 hourly statepoints– 15,526 complete MPACT/CTF converged iterations– 13.3 days continuous wall time– 1.3 million core-hours – ~6 mins per statepoint

OLCF’s TITAN Supercomputer at Oak Ridge National Laboratory

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Application - AP1000 ® PWR Advanced Core Analysis– Advanced core configuration for optimum fuel cost

and equilibrium cycle transition –Rodded depletion with MSHIMTM operation

– Excellent benchmarking opportunity for VERA with state-of-the-art in PWR core design and operation

AP1000 Sanmen site - China

Worked performed by: F. Franceschini1 D. Salazar1 A. Godfrey2

1 Westinghouse Electric Company LLC 2 Oak Ridge National Laboratory

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AP1000® PWR Advanced First Core

MSHIM BOC Power Distribution

High-resolution VERA model of the AP1000 PWR Advanced Core

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• Low Power Physics Tests (BOC)– Key startup parameters (initial critical boron, boron worth, ITC)– Rod worth

• Rod Swap and/or DRWM• Gray rods (tungsten) predictions are key• Integral and differential rod worth

• Excellent agreement for VERA vs. CE Monte-Carlo – Confirmed Westinghouse in-house predictions (nodal diffusion)

VERA simulations supporting AP1000

MA MB

MC

MD

M1

M2

AO

S1

S2

S3

S4

-15%

-10%

-5%

0%

5%

10%

15%

SHIFT MPACT Nodal Diffusion

SHIFT VERA Nodal Diffusion

HZP Critical Boron +3ppm

-9 ppm

+18 ppm

Isothermal Temp. Coeff.

-0.3pcm/F

+0.8pcm/F

+0.3pcm/F

Delta Boron and ITC vs. KENODelta Rod Worth (%) vs. KENO

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CRUD-induced power shift (CIPS)• Deviation in axial power shape

– Cause: Boron uptake in CRUD deposits in high power density regions with subcooled boiling

– Affects fuel management and thermal margin in many plants• Power uprates will increase potential for CRUD growth

Need: Multi-physics chemistry, flow, and neutronicsmodel to predict CRUD growth

CRUD deposits

CR

UD

mas

s ba

lanc

e

Thot

Tcold

Crud deposited or released by

particle and soluble mass

transfer

Crud carried over from prior cycles,

available for release

Dissolved and particulate corrosion products circulate in

coolant

Nickel/ironreleased by corrosion

-20-15-10

-505

1015

0 5000 10000 15000 20000Cycle Burnup (MWD/MTU)

Axial

Offs

et (%

)

Measured AOPredicted AO

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Application - Perform Core Design and CIPS Analysis of a Future Core Design and Compare to Industry Risk Analysis• Project with collaboration with Duke Energy

investigating application of VERA to Catawba Unit 2, Cycle 22 (current cycle)

• Industry CIPS Risk Analysis Follows EPRI Guidelines – Does not directly assess impact on CIPS on key parameters - axial offset, shutdown margin, etc.

• CASL simulation first of a kind direct analysis of CIPS axial offset for three core designs –Explicitly including the feedback of of boron on power distribution and calculating A/O

• Shows more aggressive core designs may be acceptable

Worked performed by: T. Lange1 J. Young2 B. Black2

1 University of Tennessee2 Duke Energy

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Catawba Nuclear Station

• Catawba Nuclear Station– York, SC outside of Charlotte

• Two Unit Westinghouse 4-loop PWR– Unit 1 currently on cycle 23– Unit 2 beginning cycle 22

• Duke performs all core designs and safety analyses (except LOCA)

• Cycle 22 design efficiency limited by perceived risk of CRUD-Induced Power Shift (CIPS)

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VERA Radial Boron Comparison305 cm high @ 350 EFPD

Low Risk Medium Risk High Risk

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Additional Axial Offset vs. Core Crud Boron

BOA Risk Level Max Core Crud Boron Additional AO Normalized Additional AOLow Risk 0.292 -1.75% * 0.00%Medium Risk 0.350 -2.03% 0.28%High Risk 0.410 -2.36% 0.61%

* Historical BOA risk analysis(0.3 lbm => -1.50% Additional AO)

Exceeding established CRUD boron thresholds results in marginal Additional AO when feedback incorporated

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Departure from nucleate boiling (DNB)

• Local clad surface dryout causes dramatic reduction in heat transfer during transients (e.g., overpower and loss of coolant flow)

• Current limitations:– Absence of detailed pin modeling in TH

methods results in conservative analysis– Detailed flow patterns and mixing

not explicitly modeled in single- and two-phase flow downstream of spacer grids

• Power uprates require improved quantification of margins for DNB or dryout limits

Need: High-fidelity modeling of complex flow and heat transfer for all pins in core downstream of spacer grids

Boiling Curve

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Application - Simulation of Steamline Break without offsite power (DNB Event)• Hot Zero Power Steamline Break (HZP SLB) is a cooldown event that

challenges fuel thermal limit (e.g. DNB)– Increased steamflow Reduced RCS temperature and pressure

Positive reactivity insertion Reactor core power and peaking factor increase DNB challenge

– Condition 4 event analyzed to meet Condition 2 acceptance criteria• HZP SLB cases considered in plant safety analysis

– With offsite power available and Reactor Coolant Pumps (RCPs) in operation (high flow)

– Without offsite power and natural circulation (low flow)• Problem statement:

– Which HZP SLB case is more DNB limiting, high or low flow?– Analysis of low flow case requires more effort and cost, if it is the

limiting case with respect to DNB

Work performed by: C. S. Brown1, H. Zhang2, V. Kucukboyaci3, Y. Sung3

1 North Carolina State University2 Idaho National Laboratory3 Westinghouse electric Corporation

VERA-CS 4-Loop Core Model• 56,288 channels

• 112,064 gaps • 50,952 fuel rods, 4,825 GT/IT

• ~60 axial nodes

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Study of HZP Steamline Break DNB Limiting Case

• Study based on CASL codes and technology developed and ready for application– Quasi-steady state VERA-CS

(MPACT/CTF) for high resolution full core modeling and simulation

– Core inlet temperature and flow distributions based on CFD modeling and simulation

– Sensitivity and Uncertainty quantification for limiting case determination (59-case study)

• VERA-CS coupled code predictions confirmed high flow case is more DNB limiting than low flow case

High Flow – VERA Pin Powers High Flow – VERA Core Tcoolant

CFD Inlet Temperature & FlowSTAR-CCM+ Vessel Modeling

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SLB Hot Channel Parameter Comparisons

VERA-CS based process demonstrated for future plant specific application

High flow case is more DNB limitingthan low flow case due to higher heat flux

Parameter High-Flow Low-FlowW-3 DNBR (Wilks 95/95) 3.42 4.12

DNB Limiting Elevation (cm) 45.9 30.5

Max. Pin Linear Power (W/cm) 264.3 178.5

Heat Flux (W/m2) 801.4 558.7Equilibrium Quality -0.047 -0.114Mass Flux (kg/m2/s) 4529.1 466.9

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Pellet-Clad Interaction (PCI)• PCI failure potential limits reactor performance associated with power uprates, higher burnup, fuel rod

manufacturing quality and operating flexibility during power changes• Requires new 3D multi-physics simulation capability to reduce uncertainties in assessing PCI failure

conditions during normal operation and in the presence of anomalies

Need: 3D fuel performance model to assess complex, coupled physics and irregular geometries responsible for PCI fuel failures

PCI is controlled by local effects

PCI is possible in many rods and

assemblies

PCI has system wide influence

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Cycle 9

Cycle 12

Application – Braidwood C10 & C13 PCI Failure Analysis

• Make use of the characteristics of Cycle 10 failed rod (M12S-B06), which exhibited MPS-PCI class of failure, to characterize or predict Cycle 13 failure (U22S-D03)

• Determine the minimum pellet defect (MPS) size required to exceed the stress failure threshold limit

Rod Characteristics M12S-B06 U22S-D03 UnitsFluence 5.62e25 5.68e25 n/m2

Plenum Pressure 7.57 7.00 MPaPellet-Clad Gap 17.8 17.8 micron

Rod Average Burnup 30.6 30.4 MWd/tUPeak Stress Axial

Location1.3 .8 m

Cycle 10

Cycle 13

Worked performed by: N. Capps1 J. Rashid1 B. Wirth2

1 Structural Integrity Associates2 University of Tennessee

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-120

-100

-80

-60

-40

-20

0

20

40

0 50 100 150 200 250 300 350 400 450 500

Cri

tica

l B

oro

n D

iffe

ren

ce (

pp

m)

Cycle Burnup (EFPD)

Cycle 1

Cycle 2

Cycle 3

Cycle 4

Cycle 5

Cycle 6

Cycle 7

Cycle 8

Cycle 9

Cycle 10

Cycle 11

Cycle 12

Trend

BISON Fuel Failure Analysis Methodology

Model Multiple Cycles

Run Bison for Every Pin

Select Pins of Interest

• A methodology for analysis of PCI challenge problem with VERA has been implemented and demonstrated:– Perform 2D (R-Z) fuel performance simulations to screen ever rod in the core for PCI

indicators – Selected pins of interest and perform 3-D (R-θ-Z) or 2-D (R-θ) Local Effects Analysis– Assess PCI risk based on stress failure thresholds (determined from 3-D or 2-D)

Local EffectsAnalysis

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Stress Analysis of C13 U22S-D03 Failed Rod

PCI 60 mil MPS2 3 4 5 2 3 4 5

C 306 316 315 412 418 418D 317 307* N/A 309 406 418* N/A 416E 308 305 306 400 406 420

PCI 60 mil MPS2 3 4 5 2 3 4 5

C 407 416 416 495 480 498D 413 415* N/A 403 480 494* N/A 493E 412 412 410 483 482 493

2-D R-θ stress results in MPa

3-D R-θ stress results in MPa

• The surrounding rods have similar calculated stresses and did not fail, therefore, an external factor must have contributed to the failure

Cycle 10

Cycle 13

Cycle 10

Cycle 13

Clad stress is related to the fuel centerline temperature which is directly related to the startup power history

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Strategic Applications of VERA• Commercial power industry

– NSSS and Fuel Vendors: new plant and fuel design– Owner / Operators: independent evaluation, margin enhancement, issue evaluation, backstop for licensing– EPRI: uprate, aging and issue evaluations– Consultants and support industry: tools for utility support

• US Naval reactors has requested a copy of VERA• Research and academia

– VERA is being deployed as an education tool for new engineers (currently 6 universities are using)– VERA is being adopted by universities as a research tool– Potential experimental collaboration with research reactors– VERA’s framework can support evaluation of certain Accident tolerant fuel concepts, including Mo clad

• NRC has expressed interest in VERA and is initiating a Test Stand

CASL Tools Can Support Delivering the Nuclear PromiseThrough Improved Fuel Performance & Reduced Fuel Cost

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www.casl.gov