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Coupled Codes for Multi-PhysicsNuclear Reactor Simulation

T.J. Downar

Professor

Purdue University

July 20, 2006

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Outline

Background on Coupled Codes Current Generation Methods (NRC):

e.g. RELAP5-TRACE/PARCS + TRITON

Next Generation Methods (DOE):e.g. STAR-CD/DECART

Next Next Generation Methods

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Research GroupYunlin Xu, Post-Doc

Mikhail Kalugin, Post-DocJustin Thomas, PhD Student

Shaun Clarke, PhD Student

Volkan Seker, PhD StudentEmily Decaret, MS Student

Andrew Ward, MS Student

Jere Jenkins, MS Student

Brendan Kochunas, Undergrad

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Research Group Focus

U.S. NRC Reactor Physics Methods Development

(LWR MOX, ESBWR, ACR-700, PBMR, PHWR) Coupled Neutronics/Thermal-Hydraulic Code Development for

Reactor Transient Simulation

U.S. DOE Coupled CFD/MOC (INERI/EPRI w/ ANL) Advanced Fuel Cycle Initiative (UNERI) Resonance Self-Shielding Methods (NEER w/ ORNL)

Other Homeland Security (DHS w/ ORNL) Power Grid (NSF w/ CRI)

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PARCS

A Multidimensional Multigroup ReactorKinetics Code Based on the Nonlinear

Nodal Method

Prepared for Release of PARCS / NRC-V2.7

Thomas J. Downar

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Coupled Neutron / Nuclide andTemperature/ Fluid Field Equations

Nuclide Depletion Equation (Bateman)

)()()()()(

tNtNtNdt

tdNBBcCAA

a

AA +++=

''),',',()',',(

),,(4

1),,,(),(),,,(

1

' '

+

=++

ddEtErEEr

tErStErErtErtv

E

s

ft

Neutron Transport Equation (Boltzmann)

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Solving the Core Neutronics Problem

ConventionalApproach (TRITON/PARCS)

TRITON: Lattice Calculation Involving Homogenization and

Group Condensation to Generate Few Group Xsecs PARCS: Diffusion (P1) or SP3 Approximation for Core

Calculation

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Core Neutronics Modeling

Fuel

Assembly

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Transverse Integrated Equations

3D Neutron Diffusion Equation

kkf

eff

kkkkkkkz

Dzy

Dyx

Dx

,1

=+

+

+

Nodal Methods for Solving

Reactor Neutronics

( ) ( )),(,)1,(,

,

),(,),1(,

,

),(),(,

),(),(),(),(

111jiyjiy

jy

jixjix

ix

jijif

eff

jijijiji

JJh

JJhk

zD

z

=

+

++

+ +

=1 1

),,(1

,,

),(

i

i

j

j

x

x

y

yjyix

ji dydxzyxhh

dyzyx

x

D

h

Jj

ji

y

yxxjy

jix +

=

=

1

),,(1

,

),(,

Where,

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Nodal Diffusion Methods

The three, 1-D homogeneous equations are coupledthrough a transverse leakage source term.

Nodal methods differ generally by the expansion

used for the one-dimensional flux: Analytic Nodal Method (ANM): analytic solution of the 1-Dflux

Nodal Expansion Method (NEM): polynomial expansion (4th

order) of the 1-D flux

Various innovative coarse mesh accelerationmethods (CMFD) are used to reduce thecomputational time

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PARCS Nodal NeutronicsMethods: Solution Kernels

Diffusion/SP3

MGFDFMFD

Diffusion2GFDCMFDCylindrical

3D

DiffusionMGnodalTPEN

Diffusion2GFDCMFDHexagonal

3D

SP3MGnodalNEMMG

SP3MGFDFMFD

Diffusion2GnodalANM

Diffusion2GFDCMFD

Cartesian

3D

Angle

Treatment

Energy

Treatment

Solution

Method

Kernel

Name

Geometry

Type

CMFD = Coarse Mesh Finite DifferenceANM = Analytic Nodal Method

FMFD = Fine Mesh Finite DifferenceNEMMG = Nodal Expansion Method

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Dehomogenization: Reconstructing LocalInformation from Nodal Solution

Where j represents all nodes involves for Ith LPRM,

det

, , ,( , , ) ( , ) ( , ) ( , )global local

I g j g j g j

g j I

LPRM x y z x y x y x y

=

PARCS Detector Model

det

, ,,local

g j g j

,

global

g jWill be obtained from flux reconstruction.

are provided from lattice calculation.

PARCS utilizes Analytic Function Expansion Nodal Method (AFEN) for fluxreconstruction, which involves analytic solution for 2-D diffusion solutionin homogeneous media.

Similar reconstruction is used to recover pin powers from nodal solution

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U.S. NRC Coupled Code System

Lattice Code

TRITON(ORNL)

Cross

Section

Library

(PMAX)

Neutron Flux

Solver

(PARCS)

Depletion

Module

T/H code

TRACE

(RELAP5)

GENPMAXS

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Coupling of Neutronics andThermal-Hydraulics

( )22

2

),,,,( DmDm

SbSb

DmDm

TmTm

TfTf

SbDmTmTfcrr

+

+

+

+

++=

Partials obtained by piecewise linearinterpolation

Burnup and burnup history dependence

)2,1,,()2,1,,(

)2,1,,()2,1,,(

)2,1,,()2,1,(

2

2

2

2

HISHISBUDmDmDm

HISHISBUDmDmDm

HISHISBUTmTmTm

HISHISBUTfTfTf

HISHISBUSbSbSbHISHISBUcrcr

=

=

=

=

=

=

Neutron Cross Section Model

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History of TRACE/RELAP5/PARCS Coupling

1998 PARCS / GI / RELAP Parallel Virtual Machine

1999 PARCS / GI / TRAC-M

2000 Merging GI into PARCS (Memory Copy

instead of PVM)

Automatic Mapping forPARCS / TRAC-M

2004 Merging PARCS into

TRACE Static Linking Library No PVM

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TRACE/PARCS GUI

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NRC Coupled Code Assessment

Pressurized Water Reactor Main Steam Line Break (i.e. plant uprates) LOCA (e.g. David Besse)

Boiling Water Reactor Peach Bottom Depletion/Turbine Trip Test Ringhalls Flow Instability

Next Generation Reactors ESBWR Advanced CANDU (ACR-700) PBMR

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Nuclear Reactor Transient Analysis:Main Steam Line Break

PWR Plant

Schematic

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Main Steam Line Break Transient:Core Coolant Temperature

220230

240

250

260270

280

290

300

0 20 40 60 80 100

Time [sec]

Temperature

[oC]

Intact Loop

Broken Loop

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MSLB: Reactivity Components

-8.0

-6.0

-4.0

-2.0

0.0

2.0

4.0

6.0

0 20 40 60 80 100

Tim e (s)

Reactivity($)

TotalTotal

DopplerDoppler

Control RodControl Rod

DensityDensity

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Main Steam Line Break Analysis:Core Average Power (Point Kinetics)

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Core Neutronics Model

13

4

2

65

8

79

10

11

1218

17

16

15

14

13

19

neutronic fuelassembly

reflectorassembly

stuck rodlocation

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Spatial Kinetics Analysis of MainSteam Line Break

Initial Steady-State

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MSLB Transient Analysis

Core Average Power Core Radial Power

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BWR Assessment I:Peach Bottom Turbine Trip Transient

Sudden Closure of the Turbine Stop Valve (TSV)

The Pressure Oscillation Generated in the Main Steam PipingPropagates With Relatively Little Attenuation Into the Reactor

Core

The Core Pressure Induced Oscillations Result in DramaticChanges in the Core Void Distribution and Fluid Flow

The Magnitude of the Neutron Flux/Power Increase is StronglyAffected by the Initial Rate of Pressure Rise Caused by thePressure Oscillation and has a Strong Spatial Variation

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Pressure Wave After Turbine Trip

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Core Void / Power During Transient

Core Average Void Fraction

PBTT Exercise 3

0.28

0.29

0.30

0.31

0.32

0.33

0.0 0.5 1.0 1.5 2.0

Time (s)

VoidFraction

Total Power

PBTT Exercise 3

0

50

100

150

200

250

300

0 0.2 0.4 0.6 0.8 1 1.2 1.4

Time (s)

Power(%)

TRACM/PARC S

Measurement

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BWR Assessment II: RinghallsOECD/NEA Stability Benchmark

0.490.630.500.71410477.710

0.540.990.560.80369472.69

Reg. Freq

(hz)

Regional

D.R.

Global Freq

(hz)

Global

D.R.

Flow

Kg/s

Power

%

Case

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TRACE/PARCS CPU Requirements*vrs Number of TRACE CHANs

0

2

4

6

8

10

12

204 648Number of TRACE Channels

CPU

time(hours)

TRACE

PARCS

486

1 2 3 4 5 6 7 8 9 10 1 1 1 2 1 3 14 1 5 1 6 17 1 8 1 9 20 2 1 2 2 23 2 4 2 5 26 2 7 28 29 3 0

1 1 1 1 4 5 6

2 7 8 8 10 11 12 13 14 : 8x8

3 15 16 17 18 19 19 21 21 23 24 25 26 27 28 : SEAV

4 29 30 31 32 33 34 34 36 37 37 39 40 41 42 43 44

5 45 46 47 48 49 33 51 34 53 54 55 56 57 58 59 60 43 44

6 45 64 65 66 49 68 69 70 51 72 53 74 75 76 57 78 78 80 81 52

7 45 64 85 86 87 68 89 90 91 92 93 94 95 96 75 98 99 1 00 101 102 8 1 104

8 1 05 1 06 8 5 1 08 8 7 1 1 0 1 11 1 12 8 9 9 2 9 1 9 4 1 17 1 18 9 5 1 20 1 20 1 22 9 9 1 2 4 1 01 1 26 1 27 1 28

9 1 29 1 30 1 31 1 32 1 33 1 34 1 1 1 1 3 6 1 3 7 1 3 8 1 3 9 9 4 1 41 1 42 1 17 1 44 1 45 1 4 6 1 4 7 1 4 8 1 4 9 1 5 0 1 5 1 1 5 2 1 5 3 1 5 4

10 155 156 157 158 134 133 161 162 163 137 165 139 167 141 169 144 171 145 147 149 175 176 177 178 179 180

11 181 182 183 184 185 186 187 188 189 190 191 192 193 194 195 196 197 198 199 200 201 202 203 204 179 206

12 2 07 2 0 8 2 0 9 2 0 9 2 1 1 1 8 4 1 8 6 2 1 4 1 8 8 2 1 6 2 1 7 2 1 8 1 9 1 2 2 0 1 9 3 2 22 1 95 2 24 2 2 5 2 2 6 1 9 9 2 2 8 2 0 1 2 3 0 2 0 3 2 0 4 2 3 3 2 34

13 2 35 2 36 2 3 7 2 3 8 2 3 9 1 8 4 2 4 1 2 4 2 2 4 3 2 4 3 2 4 5 2 4 6 2 1 7 2 4 8 2 4 9 2 2 0 2 51 2 52 2 53 2 5 4 2 5 4 2 5 6 2 5 7 2 5 8 2 5 9 2 3 0 2 6 1 2 6 2 2 63 2 64

14 2 65 2 66 2 6 7 2 6 8 2 6 8 2 7 0 2 7 1 2 7 2 2 7 3 2 4 3 2 7 5 2 4 5 2 7 7 2 7 8 2 7 9 2 5 1 2 81 2 82 2 83 2 8 4 2 5 6 2 8 6 2 8 7 2 8 8 2 8 9 2 8 9 2 9 1 2 9 2 2 63 2 94

15 2 65 2 96 2 6 7 2 9 8 2 9 9 2 9 9 3 0 1 3 0 1 3 0 3 3 0 4 3 0 5 3 0 6 3 0 7 3 0 8 2 7 8 3 1 0 3 11 3 11 2 82 3 1 4 3 1 5 3 1 6 3 1 7 3 1 8 3 1 9 3 2 0 2 9 2 3 2 2 2 63 3 24

16 2 65 3 26 3 2 7 2 9 8 3 2 9 3 3 0 3 3 1 3 3 2 3 3 3 3 3 4 3 3 5 3 0 6 3 0 7 3 0 7 3 3 9 3 4 0 3 41 3 11 3 43 3 4 4 3 4 5 3 4 6 3 4 7 3 4 7 3 4 9 3 1 9 3 5 1 3 5 1 3 53 3 24

17 3 55 3 26 3 5 7 3 5 8 3 5 9 3 6 0 3 6 1 3 6 2 3 6 3 3 6 4 3 6 5 3 6 6 3 6 7 3 6 8 3 6 9 3 3 9 3 71 3 72 3 73 3 7 4 3 7 5 3 7 6 3 7 7 3 7 8 3 7 9 3 8 0 3 8 1 3 8 2 3 83 3 24

18 3 85 3 26 3 8 7 3 8 8 3 8 9 3 9 0 3 9 1 3 6 1 3 6 4 3 6 3 3 9 5 3 6 5 3 9 7 3 9 8 3 9 9 4 0 0 4 01 4 02 4 03 3 7 3 4 0 5 3 7 5 4 0 7 4 0 8 4 0 9 4 1 0 3 8 0 4 1 2 4 13 4 14

19 4 15 4 1 6 4 1 7 4 1 8 3 8 9 4 2 0 4 2 1 4 2 2 4 2 3 3 9 5 4 2 5 4 2 6 4 2 7 3 9 8 4 2 9 4 00 4 31 4 02 4 3 3 4 3 4 4 3 5 4 3 6 4 3 7 4 3 8 4 3 9 4 3 9 4 4 1 4 42

20 443 444 417 446 447 420 449 450 451 452 453 426 455 456 457 458 459 460 434 462 436 464 437 466 467 468

21 469 444 417 472 446 474 475 476 477 478 479 480 481 482 456 457 485 486 487 488 489 490 491 492 467 494

22 495 496 497 498 499 475 501 502 503 477 479 506 507 481 509 510 511 485 513 487 490 516 517 518 519 494

23 521 522 523 498 525 526 527 528 528 530 531 506 509 534 535 510 537 538 539 540 541 542 519 544

24 545 546 547 548 549 526 551 552 553 530 555 556 557 534 559 560 561 540 563 542 565 544

25 567 546 547 570 570 572 573 552 575 576 577 578 579 560 561 582 563 584 565 586

26 567 588 589 590 591 572 593 594 595 576 597 598 599 582 601 602 603 586

27 567 588 607 608 609 594 611 611 613 597 597 599 617 618 619 620

28 621 622 623 624 625 626 627 627 629 630 631 632 633 620

29 635 636 637 638 639 640 640 642

30 643 644 645 646 646 646

TRACE/PARCS Problem size

Core TH ->684 x 27= 18468 TRACE cells

Core Neutronics->684 x 27=18468 PARCS nodes

~2 hours on 2 GHz machine for initialization

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Advanced Reactors

Recent emphasis has been on addressingGENIII+/GENIV licensing concerns:

ESBWR

ACR-700

HTR

Neutronics methods for MOX fuel analysis as part ofthe LWR PU Disposition program

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Advanced CANDU Reactor

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THE OECD/NEA PBMR COUPLED KINETICS CORE THERMALTHE OECD/NEA PBMR COUPLED KINETICS CORE THERMAL--HYDRAULICS BENCHMARK TEST PROBLEMSHYDRAULICS BENCHMARK TEST PROBLEMS

Frederik Reitsma PBMR Ltd, South Africa

Gerhard Strydom 1, Han de Haas 2, Kostadin Ivanov 3, Bismark Tyobeka 3,Ramatsemela Mphahlele 3*, Tom Downar 4, Volkan Seker 4, Hans D Gougar

5, D F Da Cruz 2

1 Reactor Analysis Group, PBMR (Pty) Ltd, South Africa 2 Fuel, Actinides & IsotopesGroup, NRG, The Netherlands3 Nuclear Engineering Program, Penn State University, USA 4 Nuclear EngineeringDepartment, Purdue University, USA5

Fission & Fusion Systems, INEEL, USA*

National Nuclear Regulator,Centurion, South Africa

The Second Topical Meeting on High Temperature Reactor 2004September 22-24, 2004

Beijing, China

PEBBLE BED MODULAR REACTORPEBBLE BED MODULAR REACTOR

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Goal

Develop a Numerical Nuclear Reactorwhich performs detailed, first-principle based simulation of the multi-physics phenomenaoccurring in nuclear reactors

Objectives

Provide a fully integrated, high-fidelity simulation capability that canbe used for the analysis of light water reactors and advanced

reactors

Exploit high performance computers and elaborated models tocomplement experimental verification when designing a newsystem

Next Generation Coupled Code Methods

The Numerical Nuclear Reactor(DOE INERI 2001-04)

(EPRI/DOE 2005-2007)

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Numerical Reactor Elements and

Participants

Fuel Rod Config. Loading Pattern Core Geometry

Channel Geometry

Inlet Flow Cond. Water/Fuel PropertyAssembly Structure Core Structure

45 Group

Cross Section

Library

Fuel

Performance

Code

(Anatech)

Pin-wisePower

Distribution

Pin-wise

Fuel Temp. &Fluid Force

Fine Mesh CFD

w/ConjugateHeat Transfer

Monte Carlo

w/Temp. Feedback

FuelDimension

Continuous

Xsec Library

Direct 3D

Whole Core

Transport

KAERI /Purdue

ANL

Purdue

Mechanistic and Detailed T/HCalculation

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Ray Tracing and Method of Characteristics

fewmm

in(0);)()(

==+ q

d

d

min

m

out Flat Source Region ( )ssminmout eqe

+= 1

Assume Constant Xsecand Flat Sourcewithin a Micro Region

(s = segment length)

+

=

q

s

mout

min

m

=m

mm Scalar Flux

Segment

AverageAngularFlux

Neutron Balance along Ray

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Ray Tracing Computational Rqmts (2D)

45Number of Energy Groups

~120000

Total Number of Rays

12Number of Planes

~9GBytes

Memory Required

~24

hours

Run Time (1 CPU)

~10Number of Ray Tracing Sweeps

500000Total Number of Regions

~10000Number of Cells

~50Number of Regions per Cell

0.2 mmRay Spacing

8/4# of Azimuthal/Polar Angles(per 90 Degrees)

ValueParameter

Typical Ray Tracing Parameters and

Execut ion Requirements for a Quarter

Core Problem

+

=

l

g

g

g

l

ggmin

gmout

sqs

sinexp1

sinexp,,

n2

1

)sin( l

s

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Planar MOC Solution Based 3-D CMFD

Intranodal Axial FluxShape by NEM/MOC

Ray Tracing

Global 3-D CMFDProblem

Axial Leakageas Source

Flux

z

Local 2-D MOCProblems

DifferentComposition and

Temperature

Cell Homogenized Cross Sections& Radial Cell Coupling Coefficients

Cell Average Flux& Axial Leakage

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Relative Axially Integrated Pin Power Difference between PARCS and DeCART

(PARCS - DeCART) / DeCART x 100%

PWR Comparison of

PARCS vs. DeCART (ARO-HZP)

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BWR Applications:

2x2 Assembly Benchmark

ROD H9

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Comparison w/ Monte Carlo

Radial Division

AzimuthalDivision 1 2 3

1

0.99

(-0.32%)

1.05

(0.25%)

1.11

(-0.87%)

21.00

(-0.73%)1.06

(0.89%)1.13

(-0.42%)

30.93

(0.81%)0.94

(0.91%)0.97

(-0.24%)

40.93

(-0.01%)0.93

(0.41%)0.95

(-0.50%)

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Example of CRUD on BWR Fuel Rods

Proceedings of the 2004 International Meeting on LWR Fuel Performance,

Orlando, Florida, September 19-22, 2004,

Paper 1016.Fuel Failures During Cycle 11 at River Bend, Edward Ruzauskas,

AREVA, Framatome ANP, Inc., David L Smith, Entergy Operations

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Thermal-Hydraulics: CFD Code STAR-CD Commercial code developed by CD-Adapco.

Finite-volume based code for the solution ofpressure, 3-D momentum, enthalpy, andturbulence equations over an arbitrary mesh.

Capabilities required for this project: Conjugate heat transfer Transient analysis Multi-phase models

Moving boundary and FSI Parallel computing

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CFD Modeling of a Fuel Channel

Total 192000 cells

Inlet

Outlet

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Catawba PWR Results

Coolant temperaturesat core outlet

Cladding outer surfacetemperatures at assembly interface

Twall-Tsat(C)

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STAR-CD 2-Phase Flow Model

Regime variablewithin liquid-continuous region.

Used to calculateeffective lengthscales and dragcoefficient formixed Taylor and

trailing bubbles.

Void

slug

Trailing

bubbles

Taylor

bubbles

Two-Phase CFD Simulations

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Two Phase CFD Simulations

OECD-NEA/US-NRC NUPEC BFBT Benchmark Benchmark based on NUPEC BWR Full-size Fine-

mesh Bundle Tests (BFBT)

Full-height, heated 8 pin by 8 pin assembly

tests Distributed radial profile

Temperature, pressure and detailed voiddistribution measurements

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

Area Averaged Void Fraction

NormalizedHeight

1.15 1.3 1 .15 1.3 1 .3 1.15 1 .3 1.15

1.3 0.45 0.89 0.89 0.89 0.45 1.15 1.3

1.15 0.89 0.89 0.89 0.89 0.89 0.45 1.15

1.3 0.89 0.89 0 0 0.89 0.89 1.15

1.3 0.89 0.89 0 0 0.89 0.89 1.15

1.15 0.45 0.89 0.89 0.89 0.89 0.45 1.15

1.3 1.15 0.45 0.89 0.89 0.45 1.15 1.3

1.15 1. 3 1.15 1.15 1.15 1.15 1.3 1.15

RADIAL POWER

DISTRIBUTION

Pressure

Temperature

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Summary of 2-phase CFD Research

Assembly level two-phase boiling with STAR-CD bubbly flow model (Completed)

Extended flow regime models in STAR-CD

(Ongoing) Experimental Validation of the STAR-CD two-

phase boiling models (Ongoing)

Coupled Code Applications under BWROperating Conditions (Initiated)

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Coupling Mechanics: Interface Design

One Time Overhead only

Transferred Periodically

Socket communication

File I/O

Zone-wise

temperature &

density

STAR-CD DeCART

User Input

Cell-wisetemperature &

densityCFD

InterfaceNeutronics

Geometry

decomposition

Zone-wise heat

generationCell-wise heat

generation

Typical

DeCART

pin mesh

Typical

STAR-CD

pin mesh

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Convergence of Coupled Calculations

Two important convergence criteria tracked:

DeCART side: change in the power distribution

between consecutive data exchanges:

STAR-CD side: fuel enthalpy residual, computed internally each CFD iteration

=

k

izone

k

izone

k

izone

izone q

qqq

1

max

1.0E-06

1.0E-04

1.0E-02

1.0E+00

1.0E+02

1.0E+04

1.0E+06

1.0E+08

0 100 200 300 400

STAR-CD Iteration

FuelEnthalpyResidua

l

convergence criterion

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3 x 3 Pin Array Model: Results

Midplane Temperature Distribution

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Test Problem Timing STAR-CD:

24 processors ~207,000 mesh/processor Manual decomposition (4 x 6)

DeCART:

12 processors

~13,000 mesh/processor

0

50

100

150

200

250

300

350

400

450

500

1 2 3 4 5 6 7 8 9

Data exchange

E

lapsedTime(s)

DeCART

STAR-CD

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Coupled DeCART/STAR-CD

3D Model of Small PWR DeCART discretization:

(1/4 core)

3 million flat fluxregions in DeCART

45 energy groups

8 azimuthal & 4 polarangles in 90.

70 million cells in STAR-CD (uses 1/8 coresegment). Fuel Guide tube

Gadolinia Absorbers

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Small LWR Core Calculation

Coupled DeCART/STAR-CD calculations have beenperformed on the ANL Beowulf clusterjazz:

24 processors for DeCART (10 million flat fluxregions)

104 processors for STAR-CD (70 million CFD cells).

1 processor for the external interface.

Steady-state calculations required ~5-6 hours (not yetoptimized)

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Summary of NNR Research

The Numerical Nuclear Reactor (NNR) provides an integrated, highfidelity analysis capability for PWR applications Verification and validation of phenomenological modules was successful Computational burden is high and needs to be reduced

Parallel computing More sophisticated coupling: Matrix-Free Newton Krylov

Extensions of NNR models and methods are now underway forapplication to BWRs

The Numerical Nuclear Reactor code system forms a foundation foradding important phenomenology to the analytical tool

EPRI currently supporting application to important practicalissues for LWR operation (e.g. crud deposition analysis)

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The Next Next Generation*

Neutronics 3D Transport (Kevin Clarno et al)

3D MCNP + Massively Parallel computing

Thermal-Hydraulics Interfacial Area Transport (e.g. explicit modeling

of bubble-bubble interactions)

The Grandest Nuclear Challenge

Improved Nuclear Fuel Performance Modeling Coupling to of Neutronics/TH to Fuel Performance

Code

* >3-5 years

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Thanks for you attention!

Questions?