HTGR fuel compact burnup example
Download the particle distribution file: File:Part2.inp.tgz
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set title "HTGR Depletion Benchmark, Prismatic Fuel Model"
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% Geometry:
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% --- Coated particle:
particle 1
fuel 0.0250
buffer 0.0340
PyC 0.0380
SiC 0.0415
PyC 0.0455
void
% --- Universe surrounding particles:
surf 10 inf
cell 10 6 matrix -10
% --- Read particles:
pbed 10 6 "part2.inp"
% --- Compact
surf 1 cyl 0.0 0.0 0.62250
surf 2 cyl 0.0 0.0 0.62350
cell 1 2 fill 10 -1
cell 2 2 helium 1 -2
cell 3 2 matrix 2
% Coolant channel:
surf 3 cyl 0.0 0.0 0.79400
cell 4 3 helium -3
cell 5 3 matrix 3
% --- Empry lattice position:
surf 5 inf
cell 6 4 matrix -5
% --- Compact lattice
lat 20 2 0.0 0.0 3 3 1.880
4 2 2
2 3 2
2 2 4
surf 7 hexxprism 0.0 0.0 1.880 0.0 4.93
cell 7 0 fill 20 -7
cell 8 0 outside 7
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% Material data:
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% -- Fuel:
mat fuel -10.4 burn 1
92238.03c 2.12877E-02
92235.03c 1.92585E-03
8016.03c 4.64272E-02
5010.03c 1.14694E-07
5011.03c 4.64570E-07
% --- Carbon buffer layer:
mat buffer -1.05 moder grph 6000
6000.03c 5.26449E-02
% --- Pyrolytic carbon layer:
mat PyC -1.90 moder grph 6000
6000.03c 9.52621E-02
% --- Silicon carbide layer:
mat SiC -3.18
14000.03c 4.77240E-02
6000.03c 4.77240E-02
% --- Graphite matrix:
mat matrix -1.75 moder grph 6000 burn 1 vol 47.045
6000.03c 8.77414E-02
5010.03c 9.64977E-09
5011.03c 3.90864E-08
mat helium 2.65156E-05
2004.03c 2.65156E-05
% --- Thermal scattering data:
therm grph gre7.00t
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% Calculation parameters:
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% --- Geometry plot:
plot 3 500 500 2.0
% --- Libraries:
set acelib "/xs/sss_endfb7u.xsdata"
set declib "/xs/sss_endfb7.dec"
set nfylib "/xs/sss_endfb7.nfy"
% --- Boundary conditions:
set bc 3
% --- Histories:
set pop 2000 500 10
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% Detectors:
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% --- Detector materials:
mat m92238 1.0 92238.03c 1.0
mat m92235 1.0 92235.03c 1.0
% --- Energy grid structures:
ene 1 1 1E-11 0.625E-6 20.0
ene 2 3 1000 1E-11 20.0
% --- Spectral indices (in fuel):
det 1 de 1 dm fuelp1r1 dr 102 m92238
det 2 de 1 dm fuelp1r1 dr 18 m92235
det 3 dm fuelp1r1 dr 18 m92235 dr 18 m92238
det 4 dm fuelp1r1 dr 18 m92235 dr 102 m92238
% --- Spectrum:
det 10 de 2
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% Burnup calculation:
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% --- Power density
set powdens 62.0E-3
% --- Irradiation history:
dep butot
0.1
0.5
1.0
2.0
3.0
4.0
5.0
6.0
7.0
8.0
9.0
10.0
12.0
14.0
16.0
18.0
20.0
22.0
24.0
26.0
28.0
30.0
32.0
34.0
36.0
38.0
40.0
45.0
50.0
55.0
60.0
65.0
70.0
75.0
80.0
85.0
90.0
95.0
100.0
105.0
110.0
115.0
120.0
% --- Nuclide inventory:
set inventory
% Actinides:
922350
922380
942390
942400
942410
942420
952410
952440
952450
% Fission products:
360850
380900
471101
551370
541350
621490
621510
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