|
|
| Line 80: |
Line 80: |
| | % --- Neutron population and criticality cycles: | | % --- Neutron population and criticality cycles: |
| | | | |
| − | set pop 5000 100 20 | + | set pop 500000 1000 200 |
| | set opti 1 | | set opti 1 |
| | | | |
Revision as of 11:14, 25 December 2021
Domain decomposition is used for saving computer memory by spatially dividing large burnup calculation problems between multiple calculation nodes. Each node handles particle transport only within its own domain, and particles crossing the domain boundaries are sent to another node to complete the history. The implementation in Serpent relies on a collision-based approach[1], i.e. the transfer is invoked when the particle undergoes a collision in a material which belongs to another domain. The decomposition affects only burnable materials divided into depletion zones using the div card. The most typical domain decomposition case is a full-core burnup calculation, in which the fuel is decomposed into separate domains to save memory. Interaction data for the other materials is shared by all parallel tasks.
The current implementation in Serpent 2 is relatively simple, and mainly designed for reactor cores. The number of domains is determined by the number of MPI tasks, and the division using the set dd input option. There are three division types:
- simple, based on depletion zone index, MODE 1.
- geometry-based division into sectors, MODE 2.
- geometry-based division into sectors and a central zone, MODE 3.
Notes:
- The methodology was implemented in version 2.1.31, and is still under development. Any bugs or problems should be reported at the Serpent discussion forum.
- The domain decomposition methodology allows the use of the restart capability from Serpent 2.1.32 version. The multiple restart files generated within a domain decomposition simulation can be read either in a subsequent domain decomposition simulation or within an automated depletion sequence, e.g., group-constant generation (matching the number of MPI tasks).
- From Serpent 2.1.32 version and on, the associated depletion output file in a domain decomposition simulation is not limited to parent materials.
Full-core example
The method is illustrated by a full-core example, where the depletion zone division and domain decomposition is defined by input lines:
% --- Domain decomposition ----------------------------------------------------
% --- Divide fuel into depletion zones:
div fuel sep 1
% --- Domain decomposition (1 = simple, 2 = sector, 3 = sector + center)
set dd 3
% -----------------------------------------------------------------------------
When the calculation is run with 5 MPI tasks, it produces the following output:
sss2 -mpi 5 core
(...)
ividing materials into depletion zones...
Material fuel:
- 63624 cells
- 63624 depletion zones in total
Sorting lists (this may take a while)...
OK.
Counting cells...
OK.
Decomposing 63624 divided materials into 5 domains:
Domain 1: 12744 materials (20.0%)
Domain 2: 12720 materials (20.0%)
Domain 3: 12720 materials (20.0%)
Domain 4: 12720 materials (20.0%)
Domain 5: 12720 materials (20.0%)
(...)
and a geometry plot that shows the domains in different colors (switching the current setup from MODE 3, to MODE 2, to MODE 1, respectively):
(a) sector-wise division plus central zone (MODE 3), (b) sector-wise division (MODE 2), (c) index-based division (MODE 1).
complete full-core input
% -----------------------------------------------------------------------------
set title "Monte Carlo performance benchmark"
% Based on revision 1.2 (July 2011) of Monte Carlo performance benchmark:
%
% http://www.oecd-nea.org/dbprog/MonteCarloPerformanceBenchmark.htm
% -----------------------------------------------------------------------------
% --- Neutron population and criticality cycles:
set pop 500000 1000 200
set opti 1
% --- Plot:
plot 3 1000 1000 0 -187.6 187.6 -187.6 187.6
% --- Thermal flux / fission rate mesh:
mesh 2 996 904
mesh 3 996 996
% -----------------------------------------------------------------------------
% ----- Geometry --------------------------------------------------------------
% --- Cross section library file path:
set acelib "sss_jeff31u.xsdata"
% --- Boundary between hot and cold:
surf 1 pz 0.0
% -- Water layers:
cell 31 3 cold_water -1
cell 32 3 hot_water 1
% --- Fuel pin:
pin 1
fuel 0.41
clad 0.475
fill 3
% --- Guide tube:
pin 2
fill 3 0.56
clad 0.62
fill 3
% --- Pin lattice:
lat 110 1 0.0 0.0 17 17 1.26
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 1 1 1 2 1 1 2 1 1 2 1 1 1 1 1
1 1 1 2 1 1 1 1 1 1 1 1 1 2 1 1 1
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 2 1 1 2 1 1 2 1 1 2 1 1 2 1 1
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 2 1 1 2 1 1 2 1 1 2 1 1 2 1 1
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 2 1 1 2 1 1 2 1 1 2 1 1 2 1 1
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 1 2 1 1 1 1 1 1 1 1 1 2 1 1 1
1 1 1 1 1 2 1 1 2 1 1 2 1 1 1 1 1
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
% --- assembly data:
surf 1000 inf
cell 110 4 fill 110 -1000
% --- Radial reflector:
cell 51 5 lower_radial_refl -1
cell 52 5 upper_radial_refl 1
% --- Core lattice:
lat 211 1 0.0 0.0 17 17 21.42
5 5 5 5 5 4 4 4 4 4 4 4 5 5 5 5 5
5 5 5 4 4 4 4 4 4 4 4 4 4 4 5 5 5
5 5 4 4 4 4 4 4 4 4 4 4 4 4 4 5 5
5 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 5
5 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 5
4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4
4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4
4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4
4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4
4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4
4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4
4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4
5 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 5
5 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 5
5 5 4 4 4 4 4 4 4 4 4 4 4 4 4 5 5
5 5 5 4 4 4 4 4 4 4 4 4 4 4 5 5 5
5 5 5 5 5 4 4 4 4 4 4 4 5 5 5 5 5
surf 211 sqc 0.0 0.0 182.07
cell 211 210 fill 211 -211
cell 212 210 lower_radial_refl 211 -1
cell 213 210 upper_radial_refl 211 1
% --- Core geometry:
surf 11 pz -229.0
surf 12 pz -199.0
surf 13 pz -193.0
surf 14 pz -183.0
surf 15 pz 183.0
surf 16 pz 203.0
surf 17 pz 215.0
surf 18 pz 223.0
surf 20 cyl 0.0 0.0 187.6
surf 21 cyl 0.0 0.0 209.0
surf 22 cyl 0.0 0.0 229.0
surf 23 cyl 0.0 0.0 249.0
cell 1 0 bottom_plate -22 11 -12
cell 2 0 bottom_nozzle -20 12 -13
cell 3 0 bottom_fa -20 13 -14
cell 4 0 lower_radial_refl 20 -21 12 -14
cell 5 0 fill 210 -21 14 -15
cell 6 0 upper_radial_refl 20 -21 15 -17
cell 7 0 top_fa -20 15 -16
cell 8 0 top_nozzle -20 16 -17
cell 9 0 cold_water 21 -22 12 -17
cell 10 0 top_plate -22 17 -18
cell 11 0 vessel 22 -23 11 -18
cell 666 0 outside 23 11 -18
cell 667 0 outside -11
cell 668 0 outside 18
% -----------------------------------------------------------------------------
% --- Materials ---------------------------------------------------------------
mat fuel -10.062 burn 1
92234.03c 4.9476E-6
92235.03c 4.8218E-4
92236.03c 9.0402E-5
92238.03c 2.1504E-2
93237.03c 7.3733E-6
94238.03c 1.5148E-6
94239.03c 1.3955E-4
94240.03c 3.4405E-5
94241.03c 2.1439E-5
94242.03c 3.7422E-6
95241.03c 4.5041E-7
95242.03c 9.2301E-9
95243.03c 4.7878E-7
96242.03c 1.0485E-7
96243.03c 1.4268E-9
96244.03c 8.8756E-8
96245.03c 3.5285E-9
42095.03c 2.6497E-5
43099.03c 3.2772E-5
44101.03c 3.0742E-5
44103.03c 2.3505E-6
47109.03c 2.0009E-6
54135.03c 1.0801E-8
55133.03c 3.4612E-5
60143.03c 2.6078E-5
60145.03c 1.9898E-5
62147.03c 1.6128E-6
62149.03c 1.1627E-7
62150.03c 7.1727E-6
62151.03c 5.4947E-7
62152.03c 3.0221E-6
63153.03c 2.6209E-6
64155.03c 1.5369E-9
8016.03c 4.5737E-2
mat clad -5.77 rgb 100 100 100
40000.03c 1.0
mat cold_water -0.74 moder lwtr 1001 rgb 100 150 200
1001.03c 2.000
8016.03c 1.000
5010.03c 6.490E-4
5011.03c 2.689E-3
mat hot_water -0.66 moder lwtr 1001 rgb 100 150 200
1001.03c 2.000
8016.03c 1.000
5010.03c 6.490E-4
5011.03c 2.689E-3
mat vessel -7.9 rgb 50 50 50
26000.03c -96.3
28000.03c -1.0
25055.03c -1.0
42000.03c -0.6
14000.03c -0.4
24000.03c -0.25
6000.03c -0.25
29000.03c -0.2
mat lower_radial_refl -4.32 moder lwtr 1001 rgb 100 100 100
1001.03c -0.0095661
8016.03c -0.0759107
5010.03c -3.08409E-4
5011.03c -1.40499E-5
26000.03c -0.6309028
28000.03c -0.0822917
25055.03c -0.0182870
14000.03c -0.0091435
24000.03c -0.1737269
mat upper_radial_refl -4.28 moder lwtr 1001 rgb 100 100 100
1001.03c -0.0086117
8016.03c -0.0683369
5010.03c -2.77638E-5
5011.03c -1.26481E-4
26000.03c -0.6367991
28000.03c -0.0830607
25055.03c -0.0184579
14000.03c -0.0092290
24000.03c -0.1753505
mat bottom_plate -7.184 moder lwtr 1001
1001.03c -0.0011505
8016.03c -0.0091296
5010.03c -3.70915E-6
5011.03c -1.68974E-5
26000.03c -0.6828925
28000.03c -0.0890729
25055.03c -0.0197940
14000.03c -0.0098970
24000.03c -0.1880429
mat bottom_nozzle -2.53 moder lwtr 1001
1001.03c -0.0245014
8016.03c -0.1944274
5010.03c -7.89917E-5
5011.03c -3.59854E-4
26000.03c -0.5386364
28000.03c -0.0702569
25055.03c -0.0156126
14000.03c -0.0078063
24000.03c -0.1483202
mat top_nozzle -1.746 moder lwtr 1001
1001.03c -0.0358870
8016.03c -0.2847761
5010.03c -1.15699E-4
5011.03c -5.27075E-4
26000.03c -0.4682990
28000.03c -0.0610825
25055.03c -0.0135739
14000.03c -0.0067869
24000.03c -0.1289519
mat top_plate -4.28 moder lwtr 1001
1001.03c -0.0086117
8016.03c -0.0683369
5010.03c -2.77638E-5
5011.03c -1.26481E-4
26000.03c -0.6367991
28000.03c -0.0830607
25055.03c -0.0184579
14000.03c -0.0092290
24000.03c -0.1753505
mat bottom_fa -3.044 moder lwtr 1001
1001.03c -0.0162913
8016.03c -0.1292776
5010.03c -5.25228E-5
5011.03c -2.39272E-4
40000.03c -0.8541393
mat top_fa -1.762 moder lwtr 1001
1001.03c -0.0292856
8016.03c -0.2323919
5010.03c -9.44159E-5
5011.03c -4.30120E-4
40000.03c -0.7377980
% --- Thermal scattering data
therm lwtr lwj3.00t
% -----------------------------------------------------------------------------
% --- Depletion history -------------------------------------------------------
% --- Decay and fission yield libraries:
set declib "/xs/sss_jeff31.dec"
set nfylib "/xs/sss_jeff31.nfy"
% --- Irradiation cycle:
set powdens 40.0E-3
dep butot
0.10
0.50
1.00
1.50
2.00
2.50
3.00
3.50
4.00
4.50
5.00
5.50
6.00
6.50
7.00
7.50
8.00
8.50
9.00
9.50
10.00
10.50
11.00
11.50
12.00
12.50
13.00
13.50
14.00
14.50
15.00
17.50
20.00
22.50
25.00
27.50
30.00
32.50
35.00
37.50
40.00
% --- Isotope list for inventory calculation:
set inventory all
% --- Depletion output:
set depout 2
% --- Material volumes:
set mcvol 10000000000
% -----------------------------------------------------------------------------
% --- Domain decomposition ----------------------------------------------------
% --- Divide fuel into depletion zones:
div fuel sep 1
% --- Domain decomposition (1 = simple, 2 = sector, 3 = sector + center)
set dd 3
% -----------------------------------------------------------------------------
References
- ^ García, M., Leppänen, J. and Sánchez-Espinoza, V. "A Collision-based Domain Decomposition scheme for large-scale depletion with the Serpent 2 Monte Carlo code." Ann. Nucl. Energy 152 (2021) 108026.
