1 Fuel Cycle Analysis Methods for Advanced Reactor Concepts Yunlin Xu T.K. Kim D. Tinkler T.J....
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1 Fuel Cycle Analysis Methods for Advanced Reactor Concepts Yunlin Xu T.K. Kim D. Tinkler T.J. Downar Purdue University Sept. 12, 2001 The 2001 ANS International Topic Meeting on Mathematics and Computation
1 Fuel Cycle Analysis Methods for Advanced Reactor Concepts Yunlin Xu T.K. Kim D. Tinkler T.J. Downar Purdue University Sept. 12, 2001 The 2001 ANS International
1 Fuel Cycle Analysis Methods for Advanced Reactor Concepts
Yunlin Xu T.K. Kim D. Tinkler T.J. Downar Purdue University Sept.
12, 2001 The 2001 ANS International Topic Meeting on Mathematics
and Computation
Slide 2
2 Content Motivation Depletion Code system Verification
Application on SBWR Further Improvements
Slide 3
3 Motivation NERI/DOE projects at Purdue SBWR: evolutionary
Generation III+ HCBWR: revolutionary Generation IV We need a
depletion code system for Neutonic designing: Fuel cycle analysis
Safety analysis (throughout core life)
Slide 4
4 Total Height 24.6 m Internal Diameter 6.0 m Fuel Length 2.74
m Chimney Height 9 m SBWR Reactor Vessel
Slide 5
5 SBWR Containment (GE) Gravity Driven Coolant System (GDCS)
Automatic Depressurizatio n System (ADS) Passive Containment
Cooling System (PCCS) Suppression Pool
Slide 6
6 HCBWR--principle HCBWR FBR LWR
Slide 7
7 HCBWR- strategy
Slide 8
8 HCBWR- Purdue/BNL design Tight lattice hard spectrum long
cycle length Thorium as fertile negative void coefficient all
through the cycle
10 HELIOS Gadolinium pin BP1 BP2 The octant of fuel assembly
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.
Slide 11
11 GENPXS and PMAXS The PMAXS file structures GENPX read the
outputs of HELIOS and generate PMAXS files PMAXS tabulates the XSs
of the base state and the derivatives or difference of XS of the
branches and related information
Slide 12
12 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 13
13 PARCS Purdue Advanced Reactor Core Simulator A
Multidimensional Multigroup Reactor Kinetics Code Based on the
Nonlinear Nodal Method Under NRC Contract T. J. Downar etc
Slide 14
14 PARCS Validation Boiling Water Reactor OECD Peach Bottom
Turbine Trip Benchmark OECD Ringhalls Stability Benchmark (Ongoing)
Pressurized Water Reactor: Reactivity Initiated Transients (CEA,
etc.) OECD TMI Main Steam Line Break (PARCS coupled to RELAP5 and
TRAC-M)
Slide 15
15 Coupling of Code System D2NIR P2DIR DEPLETOR Depletor Input
Nuclide Field/ Burunup PARCS Neutronics Input Thermal Hydraulics
Field Thermal Hydraulics Input PDMRRDMR RELAP /TRAC Neutron Field Q
Q T,
Slide 16
16 Coupling of Code System 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 17
17 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
Slide 18
18 PARCS Cross Section 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)
Slide 19
19 Burnup and History Calculation in Depletor 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. History( moderator
density)
Slide 20
20 Multidimension linear interpolation History 1 History 2
Burnup
Slide 21
21 Reference XS and Derivatives x0x0 xrxr No branch x1x1 x0x0
xrxr One branch x i+1 x0x0 xrxr xixi x xixi x0x0 xrxr x More than
One branch for(ppm,Tf,Tm)
Slide 22
22 Reference XS and Derivatives x2x2 x0x0 xrxr x1x1 Two
branches for D More than Two branches for D x i x i+1 x i+2 x 0 x r
x x i x i+1 x 0 x r x i+2 x
Slide 23
23 Gadolinium pin BP1 BP2 The octant of fuel assembly
Verification Maximum Difference 210 -5 Comparison with HELIOS
Problem 1: Single Assembly with reflective B.C.
Slide 24
24 Verification Maximum Difference 0.3% Compared with MASTER
(KEARI) Problem 2 Checkerboard small core with vaccum B.C.
Slide 25
25 600 MWe SBWR Design Fresh fuel, 69 Once burned Fuel, 69
Twice burned Fuel, 45 Total 732 Fuel Assemblies
27 RELAP5 Model Currently uses 63 Core channels Models Vessel
Only From Cold to Hot Leg Inlet: Feedwater Tank Outlet: Turbine
Side Similar Models built for 200 and 1200 MWe
31 Depletion Results B (GWd/MT) C.R. Notches (798) Max
PeakRadial PeakAxial PeakB Average (GWd/MT)
0.0004861.7961.3071.30213.9 1.14561.7431.2801.25415.0
2.24561.7971.2711.25816.1 3.34561.8761.2631.27217.2
4.44561.9841.2621.29918.3 5.54632.1421.2961.35519.4
6.64982.3421.3321.44020.5 7.74982.3271.3501.44421.6
8.84912.1601.3531.38622.7 9.94221.8841.3441.25223.8
11.02171.8361.3251.38224.9 12.102.0691.3041.36826.0
Slide 32
32 Other Related Works VIPRE as a T/H solver for Depletion
system An EPRI code, with steady state option Used for 200MW and
1200 MW SBWR design Used for HCBWR analysis Used for Ringhalls
Benchmark problem Successfully treat the history effect
Slide 33
33 Further improvements Predictor-corrector Time integration
method Microscopic depletion?