2 Yunlin Xu T.K. Kim T.J. Downar School of Nuclear Engineering
Purdue University March 28, 2001
Slide 3
3 Content Motivation What is Depletion? Depletion code system
Verification Further improvements
Slide 4
4 Motivation Why do we need depletion code system? Basic tool
for Nuclear Reactor fuel cycle analysis NERI/DOE projects at Purdue
SBWR HCBWR Nuclear Power Reactor Analysis Economics Safety
(throughout core life)
Slide 5
5 What is Depletion? Nuclide density change in nuclear reactor
core when operated at power Related changes Nuclide density (Heavy
metal, Fission products) Cross Section Cross Section feedback Decay
Heat Reactivity economic safety Depletion code system must solve
coupled nuclide/neutron and temperature/fluid field equations
Slide 6
6 Heavy Metal Chains Arrow up :neutron capture Arrow
down:(n,2n) reaction Arrow left :electron capture Arrow right:
decay or decay for Am242 m
Slide 7
7 Equations for Depletion Nuclide depletion equation (Bateman)
B C A n, Absorb netron Neutron Transport Equation (Boltzmann)
Slide 8
8 Micro vs Macroscopic Depletion Microscopic Macroscopic
Lattice code provide Lattice code provide Solve for Nuclide Field
from the Bateman equation N/A (Nuclide density and micro changes
are combined) change depend on N i and change depend on Burnup
ComplicatedEasy to implement Smaller history effectLarger history
effect
10 HELIOS and PMAX HELIOS is a comercial (Studsvik Scandpower)
lattice physics code for solving Boltzmann equation with fine
energy group, heterogeneous, two- Dimensional models of the fuel
lattice HELIOS uses consistent fuel assembly homogenization and
energy group collapsing methods to produce few group cross sections
at all fuel assembly conditions throughout the burnup cycle. PMAX
tabulates the XSs of the base state and the derivatives or
difference of XS of the branches Gadolinium pin BP1 BP2 The octant
of fuel assembly
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11 Base state and Branches Base stateBranches 0GWD/T Fuel temp.
T f1, T f2 mod temp. T m1, T m2 Mod. den. D m1, D m2 Soluble B. ppm
1, Control rod 5GWD/T 4GWD/T 3GWD/T 1GWD/T 2GWD/T Fuel temp. T f1,
T f2 mod temp. T m1, T m2 Mod. den. D m1, D m2 Soluble B. ppm 1,
Control rod
Slide 12
12 Reactor Core Configuration Characteristics of Configuration
Heterogeneous in Radial Direction - Fuel Assemblies - Fissionable
Absorbers - Control Banks - Reflectors Homogeneous / Heterogeneous
in Axial Direction
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13 PARCS Purdue Advanced Reactor Core Simulator A
Multidimensional Multigroup Reactor Kinetics Code Based on the
Nonlinear Nodal Method Under NRC Contract Thomas J. Downar Han Gyu
Joo Douglas A. Barber Matt Miller
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14 PARCS Validation Pressurized Water Reactor: Reactivity
Initiated Transients (CEA, etc.) OECD TMI Main Steam Line Break
(PARCS coupled to RELAP5 and TRAC-M) Boiling Water Reactor OECD
Peach Bottom Turbine Trip Benchmark OECD Ringhalls Stability
Benchmark (Ongoing)
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15 PARCS The Cross Section representation used in PARCS Where r
: XS at reference state ppm : soluble boron concentration (ppm) Tf
: fuel temperature (k) Tm : moderator temperature (k) D : moderator
density (g/cc)
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16 Coupling of PARCS to TRAC-M/RELAP5 Coupling of PARCS to
DEPLETOR T/H Data Map Thermal Hydraulics Memory Structure (A) (A)
(AB) Thermal Hydraulics Input T/H Side Interface Input General
Interface Neut. Data Map Neutronics Memory Structure (AB) Memory
Structure (B) (AB) (B) Neut. Side Interface Input Neutronics Input
P2DIR DEPLETOR Memory Structure (A) (A) (AB) Depletor Input Depl.
Side Interface Input Neut. Data Map Neutronics Memory Structure (B)
(AB) (B) Neut. Side Interface Input Neutronics Input
Slide 17
17 Depletion code system based on PARCS In order to minimize
the changes to PARCS, A separate code DEPLETOR was developed The
general interface used to couple TH (RELAP5) and PARCS was used to
coupled DEPLETOR to PARCS The message transfer between PARCS and
DEPLETOR is performed using the standard message passing interface
software PVM. P2DIR, a module to communicate with DEPLETOR, was
created in PARCS (only 5 entry points in PARCS)
Slide 18
18 Algorithm for Depletion code system Read inputs Initialize
PVM Calculate XS Receive XS Send XS Neutron Flux Calc Burnup Clac
Send FluxesReceive Fluxes END EOC END PARCS DEPLETOR XS &
Derivatives Flux & XS Nodalization Exchange ID
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19 Coupling PARCS/DEPLETOR to TH EOC D2NIR(1) D2NIR(2) D2NIR(4)
D2NIR(3) DEPLETION READINP DEPLETOR INITIAL XSB y n D2NIR(2)XSB End
RELAP/TRAC R(T)DMR(1) R(T)DMR(2) R(T)DMR(3) End done y n PARCS
CHANGECOMI EOC P2DIR(3) P2DIR(4) P2DIR(2) P2DIR(1) depl PREPROC
INPUTD depl SSEIG depl extth INIT PDMR(2) PDMR(3) PDMR(1) Thconv
SCANINPUT CHANGEDIM depl y y y y y y n n n n n n P2DIR(2) End
Slide 20
20 Cross Section Model used in Depletor Interpolating XS for a
Specified burnup Using a Tabular XS Set Calculating the Burnup
Distribution. B(i) : burnup increment of ith region Bc : Core
average burnup increment G(i) : the heavy metal loading in ith
region Gc : total heavy metal loading in the core P(i) : Power in
ith region Pc : Total power in core.
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21 Cross Section Model used in Depletor Calculating XS and
Derivatives at Reference States No Branch State Case One Branch
State Case Two Branch States Case
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22 Gadolinium pin BP1 BP2 The octant of fuel assembly
Verification Problem 1: Single Assembly with reflective B.C.
Maximum Difference 210 -5 Comparison with HELIOS
Slide 23
23 Verification Problem 2 Checkerboard small core with vaccum
B.C. Maximum Difference 0.3% Compared with MASTER (KEARI)
Slide 24
24 BWR model Mapping between Neutronic and T/H model 201 101
301 202 102 302 203 103 303 Upper Plenum: 400 Lower Plenum: 100 401
099 TANK SINK Plenum to Plenum T/H model A B B A Neutronic
model
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25 Comparison between RELAP and VIPRE RELAP and TRAC are
transient codes and do not solve the steady-state
thermal-hydraulics equations We therefore examined another T/H
code, VIPRE (EPRI), which has a steady state option There are three
models in VIPRE: HEM, Drift Flux Model, and Two Fluid Model Drift
Flux Model was used for preliminary comparison RELAPVIPREDIFFERENCE
TH steps per depletion step112375-93.3%
keff1.08165021.0816311-1.9pcm fxy1.08971.0878-0.17%
fz1.80661.82000.74% Exit void Fraction Chan-10.65720.65780.06%
Chan-20.71500.71720.22% Maximum fuel Temperature (K)
Chan-12144.42153.99.5 Chan-21847.31844.8-2.5
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26 Comparison between RELAP and VIPRE
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27 Comparison between RELAP and VIPRE There is generally good
agreement between RELAP and VIPRE The only visible difference is
the fluid temperature which may be due to the sub-cooled void
model. VIPRE provides LEVY and EPRI models (The EPRI model is used
in this comparison)
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28 Further improvements VIPRE Two Fluid Model History effects
in Macroscopic X-sections Predictor-corrector Time integration
method Microscopic depletion?