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merge into: crp:rings
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crp:fom-model
crp:fullcore-long
crp:fullcore-short
crp:ecp-assembly-short
crp:zlattice
crp:rings
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pull from: crp:main
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44 commits
rings ... main

Author SHA1 Message Date
Paul Romano
db5c6bfaf8 Fix sourcepoint write in core scripts 2022-06-08 14:26:56 -05:00
Paul Romano
b47c173471 Fix --no-multipole argument for build-assembly.py 2022-06-08 12:44:21 -05:00
Paul Romano
057e6910d9
Merge pull request #18 from mit-crpg/assembly-update 
Fix the build-assembly.py script
 2022-06-08 11:25:18 -05:00
Paul Romano
ca7d63327f Add a --clone argument for build-assembly.py 2022-06-08 11:23:52 -05:00
Paul Romano
7805e62674 Fix assembly assignment in build-assembly.py 2022-06-08 11:23:49 -05:00
Paul Romano
30dfc4395e
Merge pull request #17 from mit-crpg/cleanup 
Small cleanup
 2022-06-06 22:09:05 -05:00
Paul Romano
3bdd23096a Put clone function in smr/materials.py 2022-06-06 22:07:36 -05:00
Paul Romano
f62a9c610f Add markdown description of milestone models 2022-06-06 22:01:59 -05:00
Paul Romano
c7b89db265
Merge pull request #16 from mit-crpg/fullcore-long 
Add scripts for core models used in AD-SE-08-73 milestone
 2022-06-06 21:53:22 -05:00
Paul Romano
9c4c8c87fb Only modify lattice_pitch in build-core-short/long.py scripts 2022-06-06 21:39:28 -05:00
Paul Romano
30355e218c Add script for building full length core model 2022-06-06 17:12:37 -05:00
Paul Romano
2a8764b04f Change enrichment pattern to flatten power distribution 2022-06-06 17:12:37 -05:00
Paul Romano
c9554e003a Fix volume assignment to clad/gap, use proper assemblies 2022-06-06 17:12:37 -05:00
Paul Romano
1713a4df1b Fix volumes in fullcore short model 2022-06-06 17:12:37 -05:00
Paul Romano
6abc157770 Add script for building short full core, manually change lattice_pitch 2022-06-06 17:12:37 -05:00
Paul Romano
0b3546f23a Remove assemblies with burnable absorber pins 2022-06-06 17:11:40 -05:00
Paul Romano
c375e6d9cc
Merge pull request #15 from mit-crpg/whitespace 
Whitespace improvements
 2022-06-06 17:11:09 -05:00
Paul Romano
f9624e3e2c Remove unnecessary import in surfaces.py 2022-06-06 17:10:03 -05:00
Paul Romano
4ad8676efc Only whitespace changes 2022-06-06 17:09:55 -05:00
Paul Romano
631fefebb8
Merge pull request #14 from mit-crpg/optional-differentiation 
Make differentiation of materials optional for assembly-long and core-fresh
 2022-06-06 16:59:53 -05:00
Paul Romano
1718798735 Make differentiation of materials optional for assembly-long and core-fresh 2022-06-06 16:49:18 -05:00
Paul Romano
2b8e9e9f5d Remove tallies, don't clone materials unless needed 2022-06-06 13:34:02 -05:00
Paul Romano
e53ffa6ece
Merge pull request #13 from mit-crpg/ecp-assembly-long 
Add script for assembly model used in AD-SE-08-61 milestone
 2022-06-06 13:21:41 -05:00
Paul Romano
b5b6f4611a Make sure model build doesn't fail if rings == 1 2022-06-06 13:03:35 -05:00
Paul Romano
6448eaff09 Use multipole cross sections by default 2019-12-19 11:33:13 -06:00
Paul Romano
2a24952857 Add script to build long assembly 2019-12-19 10:02:32 -06:00
Paul Romano
0e409cba87 Add script for short single assembly ExaSMR coupled runs 2019-12-19 09:09:17 -06:00
Paul Romano
5481118862 Allow configurable fuel in ring radii 2019-12-19 09:09:17 -06:00
Paul Romano
314331f275 Don't override empty guide tube universe (without spacers) 2019-12-19 09:09:17 -06:00
Paul Romano
6ae045801e Remove default temperatures on materials 2019-12-19 09:09:17 -06:00
Paul Romano
1957974073 Don't deepcopy surfaces 
PR #1393 in OpenMC changed how memoization during XML writing is
done. Specifically, the check for surfaces no longer looks at the IDs already
written.
 2019-12-19 09:09:17 -06:00
Paul Romano
6ef892a510
Merge pull request #12 from mit-crpg/temps-and-mats 
Updates to SMR model
 2019-07-24 23:08:17 -05:00
Paul Romano
6fd5188890 Remove neutron shields, which don't exist for NuScale model 2019-07-24 22:16:54 -05:00
Paul Romano
7e8feea6b1 Use SS 302 for the plenum spring, M5 for end plugs 2019-07-24 21:54:43 -05:00
Paul Romano
2d88d33730 Add ability to change reference z value 2019-07-24 21:54:12 -05:00
Paul Romano
9a24a5c4f5 Use RPV dimensions from NuScale DCA 2019-07-24 21:53:06 -05:00
Paul Romano
9ea73dbce3 Update cladding material, default temperature, water density, boron level 2019-06-19 12:50:18 -05:00
Paul Romano
b7eeb8e96e
Merge pull request #11 from mit-crpg/shallow-clone 
Shallow clone and bug fix
 2019-06-19 10:51:46 -05:00
Paul Romano
b400c1ca21 Fix axial segmentation of fuel pin 2019-06-04 22:25:45 -05:00
Paul Romano
99efb29643 Use lowercase arguments on OpenMC surfaces 2019-05-03 11:14:57 -05:00
Paul Romano
e6323e9a6f Use shallow copy of material for differentiation 2019-05-03 10:54:37 -05:00
Paul Romano
33b000696d
Merge pull request #10 from mit-crpg/detailed-model 
Update SMR model allowing subdividing fuel regions
 2018-02-28 06:51:36 -06:00
Paul Romano
f348a5ce63 Add -o option in build-core-fresh.py 2018-02-07 08:00:52 -05:00
Paul Romano
2e1761c6fb Squashed commit of the following: 
commit fea65f02928ed7b65b0aa88e16af0e9f308dc28e
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Mon Feb 5 09:20:20 2018 -0600

    Remove XML files and old models

commit 423b3ade32be620dab7ba6fe9c292a6ee4758e9e
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Mon Feb 5 09:10:54 2018 -0600

    Fix bug in ring generation

commit f5a3acd3d6d0eba7edf4b0c94e76aebdcaa2cd62
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Mon Feb 5 09:05:51 2018 -0600

    Add option to specify output directory

commit 0c4bca51b38bbb94fffe1d13157f693bee4dea03
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Tue Jan 23 14:50:09 2018 -0600

    Add plots for assembly

commit 2813d95ea694259bd50ae7fa19087f8652fc0bc5
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Sun Jan 21 16:08:45 2018 -0600

    Shared compositions when building assembly

commit 5ee50611df13f5df03144266d2c08f2890fdaf72
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Sun Jan 21 15:32:48 2018 -0600

    Fix build-core-fresh

commit e375ad22a7fb05def86664f5be8411ee822e23a2
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Fri Jan 19 15:51:51 2018 -0600

    Use pathlib in build-assembly and move assembly directory

commit 53d38a3b3711b44dd75313132c4122f2192df16c
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Fri Jan 19 14:39:09 2018 -0600

    Add an option to use depleted materials

commit 9c52f82bd0cd989b10938c85586caba331efa1f4
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Wed Jan 17 14:02:27 2018 -0600

    Add SMR assembly model

commit 82f60acfb3b8889d812e03eb4ee2007b2a5a501e
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Wed Jan 17 12:34:35 2018 -0600

    Add docstrings here and there

commit a957d442a8
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Wed Jan 17 12:18:11 2018 -0600

    Special treatment for a single radial/axial region in fuel pins

commit dac1af47ad
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Wed Jan 17 06:54:24 2018 -0600

    Add number of rings/axial segments as command-line options

commit c1082420e4
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Wed Jan 17 06:47:03 2018 -0600

    Use argparse in build-fresh.py

commit a20d15c0cb
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Wed Jan 17 06:27:01 2018 -0600

    Make number of rings configurable

commit 96b1ae8513
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Tue Jan 16 22:44:15 2018 -0600

    Generate pin universes within function

commit 0b1965e27f
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Tue Jan 16 22:28:44 2018 -0600

    Put geometry and reflector/assembly universes in functions

commit 337d4cff69
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Wed Nov 22 12:22:17 2017 -0600

    Make sure sleeve appears on fifth grid spacer

commit 46e379e169
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Tue Nov 14 14:54:14 2017 -0600

    Don't break up fuel region over multiple universes in z direction

commit 7e735bec85
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Fri Nov 10 11:49:36 2017 -0600

    Fix number of pellets in SMR

commit 186c525c8a
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Thu Nov 2 23:27:48 2017 -0500

    Add axial subdivision of fuel pins

commit 6762ab0237
Author: Paul Romano <paul.k.romano@gmail.com>
Date:   Fri Oct 20 07:54:57 2017 -0400

    Add ten rings in fuel
 2018-02-05 09:26:05 -06:00
79 changed files with 61470 additions and 6560 deletions
Download patch file Download diff file Expand all files Collapse all files

5
.gitattributes vendored
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@ -1,4 +1 @@
assembly/depleted/materials.xml filter=lfs diff=lfs merge=lfs -text
2x2-periodic/depleted/materials.xml filter=lfs diff=lfs merge=lfs -text
2x2-reflector/depleted/materials.xml filter=lfs diff=lfs merge=lfs -text
smr/depleted/materials.xml filter=lfs diff=lfs merge=lfs -text
smr/core-depleted/materials.xml filter=lfs diff=lfs merge=lfs -text

59
2x2-periodic/build-depleted.py
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@ -1,59 +0,0 @@
#!/usr/bin/env python3
import numpy as np
import opendeplete
from geometry import openmc_geometry
# FIXME: Automatically extract info needed to calculate burnable cell volumes
# Fuel rod geometric parameters
radius = 0.39218
height = 5.
# Count the number of instances for each cell and material
openmc_geometry.determine_paths(instances_only=True)
# Extract all cells filled by a fuel material
fuel_cells = openmc_geometry.get_cells_by_name(
name='enr radial 0: Fuel', case_sensitive=True)
# Assign distribmats for each material
for cell in fuel_cells:
cell.fill.volume = np.pi * radius**2 * height
cell.fill.depletable = True
cell.fill.temperature = 300.0
cell.fill = [cell.fill.clone() for i in range(cell.num_instances)]
# Set temperature for all cells
cells = openmc_geometry.get_all_cells()
for cell_id in cells:
cells[cell_id].temperature = 300.0
# Create dt vector for 1 month with 5 day timesteps
dt1 = 5*24*60*60 # 5 days
dt2 = 1.*30*24*60*60 # 1 months
N = np.floor(dt2/dt1)
dt = np.repeat([dt1], N)
# Create settings variable
settings = opendeplete.OpenMCSettings()
settings.openmc_call = "openmc"
settings.particles = 120000
settings.batches = 20
settings.inactive = 10
settings.lower_left = [-21.41728, -21.41728, +192.5]
settings.upper_right = [+21.41728, +21.41728, +197.5]
settings.entropy_dimension = [17*2, 17*2, 1]
# MeV/second cm from CASMO
settings.power = 2.337e15 * ((17.*17.*2.) / 1.5**2) * height
settings.dt_vec = dt
settings.output_dir = 'depleted'
op = opendeplete.OpenMCOperator(openmc_geometry, settings)
# Perform simulation using the MCNPX/MCNP6 algorithm
opendeplete.cecm(op)

140
2x2-periodic/build-fresh.py
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@ -1,140 +0,0 @@
#!/usr/bin/env python3
"""Creates a 2D 2x2 assembly colorset with periodic BCs."""
import os
import shutil
import copy
import numpy as np
import openmc
from geometry import beavrs, openmc_geometry
#### Query the user for options
# Query the user on whether to use multipole cross sections
multipole = input('Use multipole cross sections? (y/n): ').lower()
multipole = (multipole == 'y')
# Query the user on whether to use distribmats or distribcells
# If using distribmats, the geometry must be "differentiated" with unique
# material instances for each instance of a fuel cell
distrib = input('Use distribmat or distribcells? [mat/cell]: ').lower()
if distrib not in ['cell', 'mat']:
raise InputError('Distrib type "{}" is unsupported'.format(distrib))
#### "Differentiate" the geometry if using distribmats
if distrib == 'mat':
# Count the number of instances for each cell and material
openmc_geometry.determine_paths()
# Extract all cells filled by a fuel material
fuel_cells = openmc_geometry.get_cells_by_name(
name='enr radial 0: Fuel', case_sensitive=True)
# Assign distribmats for each material
for cell in fuel_cells:
new_materials = []
for i in range(cell.num_instances):
new_materials.append(cell.fill.clone())
# Fill cell with list of "differentiated" materials
cell.fill = new_materials
#### Create OpenMC "materials.xml" file
all_materials = openmc_geometry.get_all_materials()
materials = openmc.Materials(all_materials.values())
materials.export_to_xml()
#### Create OpenMC "geometry.xml" file
openmc_geometry.export_to_xml()
#### Create OpenMC "settings.xml" file
# Construct uniform initial source distribution over fissionable zones
lower_left = [-21.41728, -21.41728, +192.5]
upper_right = [+21.41728, +21.41728, +197.5]
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
source.space.only_fissionable = True
settings_file = openmc.Settings()
settings_file.batches = 10
settings_file.inactive = 5
settings_file.particles = 10000
settings_file.output = {'tallies': False}
settings_file.source = source
settings_file.sourcepoint_write = False
if multipole:
settings_file.temperature = {'multipole': True, 'tolerance': 1000}
settings_file.export_to_xml()
#### Create OpenMC "plots.xml" file
# Initialize the BEAVRS color mapping scheme
beavrs.write_openmc_plots()
# Create a plot colored by materials
plot = openmc.Plot()
plot.width = [21.41728*2, 21.41728*2]
plot.origin = [0., 0., 195.]
plot.color_by = 'material'
plot.filename = '2x2-periodic'
plot.pixels = [1000, 1000]
plot_file = openmc.Plots([plot])
plot_file.export_to_xml()
#### Create OpenMC "tallies.xml" file
tallies = openmc.Tallies()
# Extract all fuel materials
materials = openmc_geometry.get_materials_by_name(name='Fuel', matching=False)
# If using distribcells, create distribcell tally needed for depletion
if distrib == 'cell':
fuel_cells = openmc_geometry.get_cells_by_name(
name='enr radial 0: Fuel', case_sensitive=True, matching=False)
for cell in fuel_cells:
tally = openmc.Tally(name='depletion tally')
tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
'fission', '(n,2n)', '(n,3n)', '(n,4n)']
tally.nuclides = cell.fill.get_nuclides()
tally.filters.append(openmc.DistribcellFilter([cell.id]))
tallies.append(tally)
# If using distribmats, create material tally needed for depletion
elif distrib == 'mat':
tally = openmc.Tally(name='depletion tally')
tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
'fission', '(n,2n)', '(n,3n)', '(n,4n)']
tally.nuclides = materials[0].get_nuclides()
material_ids = [material.id for material in materials]
tally.filters.append(openmc.MaterialFilter(material_ids))
tallies.append(tally)
tallies.export_to_xml()
#### Move all XML files to 'fresh' directory
if not os.path.exists('fresh'):
os.makedirs('fresh')
shutil.move('materials.xml', 'fresh/materials.xml')
shutil.move('geometry.xml', 'fresh/geometry.xml')
shutil.move('settings.xml', 'fresh/settings.xml')
shutil.move('tallies.xml', 'fresh/tallies.xml')
shutil.move('plots.xml', 'fresh/plots.xml')

242
2x2-periodic/depleted/geometry.xml
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@ -1,242 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="10001" material="10014" name="Intermediate grid pincell radial 0: Borated Water" region="10010 -10011 10012 -10013" universe="10001" />
<cell id="10002" material="10006" name="Intermediate grid pincell radial outer: Zircaloy 4" region="~(10010 -10011 10012 -10013)" universe="10001" />
<cell id="10003" material="10014" name="Top/Bottom grid pincell radial 0: Borated Water" region="10006 -10007 10008 -10009" universe="10002" />
<cell id="10004" material="10005" name="Top/Bottom grid pincell radial outer: Inconel 718" region="~(10006 -10007 10008 -10009)" universe="10002" />
<cell id="10009" material="10014" name="Grids axial universe axial 0: Borated Water" region="-10000" universe="10005" />
<cell id="10010" material="10015" name="Grids axial universe axial 1: Water SPN" region="-10003 10000" universe="10005" />
<cell id="10011" material="10014" name="Grids axial universe axial 2: Borated Water" region="-10018 10003" universe="10005" />
<cell fill="10002" id="10012" name="Grids axial universe axial 3: Top/Bottom grid pincell" region="-10019 10018" universe="10005" />
<cell id="10013" material="10014" name="Grids axial universe axial 4: Borated Water" region="-10020 10019" universe="10005" />
<cell fill="10001" id="10014" name="Grids axial universe axial 5: Intermediate grid pincell" region="-10021 10020" universe="10005" />
<cell id="10015" material="10014" name="Grids axial universe axial 6: Borated Water" region="-10022 10021" universe="10005" />
<cell fill="10001" id="10016" name="Grids axial universe axial 7: Intermediate grid pincell" region="-10023 10022" universe="10005" />
<cell id="10017" material="10014" name="Grids axial universe axial 8: Borated Water" region="-10024 10023" universe="10005" />
<cell fill="10001" id="10018" name="Grids axial universe axial 9: Intermediate grid pincell" region="-10025 10024" universe="10005" />
<cell id="10019" material="10014" name="Grids axial universe axial 10: Borated Water" region="-10026 10025" universe="10005" />
<cell fill="10001" id="10020" name="Grids axial universe axial 11: Intermediate grid pincell" region="-10027 10026" universe="10005" />
<cell id="10021" material="10014" name="Grids axial universe axial 12: Borated Water" region="-10028 10027" universe="10005" />
<cell fill="10001" id="10022" name="Grids axial universe axial 13: Intermediate grid pincell" region="-10029 10028" universe="10005" />
<cell id="10023" material="10014" name="Grids axial universe axial 14: Borated Water" region="-10030 10029" universe="10005" />
<cell fill="10001" id="10024" name="Grids axial universe axial 15: Intermediate grid pincell" region="-10031 10030" universe="10005" />
<cell id="10025" material="10014" name="Grids axial universe axial 16: Borated Water" region="-10032 10031" universe="10005" />
<cell fill="10002" id="10026" name="Grids axial universe axial 17: Top/Bottom grid pincell" region="-10033 10032" universe="10005" />
<cell id="10027" material="10014" name="Grids axial universe axial 18: Borated Water" region="-10004 10033" universe="10005" />
<cell id="10028" material="10015" name="Grids axial universe axial 19: Water SPN" region="-10005 10004" universe="10005" />
<cell id="10029" material="10014" name="Grids axial universe axial top: Borated Water" region="10005" universe="10005" />
<cell id="10051" material="10006" name="Fuel rod lower/upper fitting radial 0: Zircaloy 4" region="-10036" universe="10007" />
<cell id="10052" material="10014" name="Fuel rod lower/upper fitting radial outer: Borated Water" region="10036" universe="10007" />
<cell id="10053" material="10017 10018 10019 10020 10021 10022 10023 10024 10025 10026 10027 10028 10029 10030 10031 10032 10033 10034 10035 10036 10037 10038 10039 10040 10041 10042 10043 10044 10045 10046 10047 10048 10049 10050 10051 10052 10053 10054 10055 10056 10057 10058 10059 10060 10061 10062 10063 10064 10065 10066 10067 10068 10069 10070 10071 10072 10073 10074 10075 10076 10077 10078 10079 10080 10081 10082 10083 10084 10085 10086 10087 10088 10089 10090 10091 10092 10093 10094 10095 10096 10097 10098 10099 10100 10101 10102 10103 10104 10105 10106 10107 10108 10109 10110 10111 10112 10113 10114 10115 10116 10117 10118 10119 10120 10121 10122 10123 10124 10125 10126 10127 10128 10129 10130 10131 10132 10133 10134 10135 10136 10137 10138 10139 10140 10141 10142 10143 10144 10145 10146 10147 10148 10149 10150 10151 10152 10153 10154 10155 10156 10157 10158 10159 10160 10161 10162 10163 10164 10165 10166 10167 10168 10169 10170 10171 10172 10173 10174 10175 10176 10177 10178 10179 10180 10181 10182 10183 10184 10185 10186 10187 10188 10189 10190 10191 10192 10193 10194 10195 10196 10197 10198 10199 10200 10201 10202 10203 10204 10205 10206 10207 10208 10209 10210 10211 10212 10213 10214 10215 10216 10217 10218 10219 10220 10221 10222 10223 10224 10225 10226 10227 10228 10229 10230 10231 10232 10233 10234 10235 10236 10237 10238 10239 10240 10241 10242 10243 10244 10245 10246 10247 10248 10249 10250 10251 10252 10253 10254 10255 10256 10257 10258 10259 10260 10261 10262 10263 10264 10265 10266 10267 10268 10269 10270 10271 10272 10273 10274 10275 10276 10277 10278 10279 10280 10281 10282 10283 10284 10285 10286 10287 10288 10289 10290 10291 10292 10293 10294 10295 10296 10297 10298 10299 10300 10301 10302 10303 10304 10305 10306 10307 10308 10309 10310 10311 10312 10313 10314 10315 10316 10317 10318 10319 10320 10321 10322 10323 10324 10325 10326 10327 10328 10329 10330 10331 10332 10333 10334 10335 10336 10337 10338 10339 10340 10341 10342 10343 10344 10345 10346 10347 10348 10349 10350 10351 10352 10353 10354 10355 10356 10357 10358 10359 10360 10361 10362 10363 10364 10365 10366 10367 10368 10369 10370 10371 10372 10373 10374 10375 10376 10377 10378 10379 10380 10381 10382 10383 10384 10385 10386 10387 10388 10389 10390 10391 10392 10393 10394 10395 10396 10397 10398 10399 10400 10401 10402 10403 10404 10405 10406 10407 10408 10409 10410 10411 10412 10413 10414 10415 10416 10417 10418 10419 10420 10421 10422 10423 10424 10425 10426 10427 10428 10429 10430 10431 10432 10433 10434 10435 10436 10437 10438 10439 10440 10441 10442 10443 10444 10445 10446 10447 10448 10449 10450 10451 10452 10453 10454 10455 10456 10457 10458 10459 10460 10461 10462 10463 10464 10465 10466 10467 10468 10469 10470 10471 10472 10473 10474 10475 10476 10477 10478 10479 10480 10481 10482 10483 10484 10485 10486 10487 10488 10489 10490 10491 10492 10493 10494 10495 10496 10497 10498 10499 10500 10501 10502 10503 10504 10505 10506 10507 10508 10509 10510 10511 10512 10513 10514 10515 10516 10517 10518 10519 10520 10521 10522 10523 10524 10525 10526 10527 10528 10529 10530 10531 10532 10533 10534 10535 10536 10537 10538 10539 10540 10541 10542 10543 10544" name="Fuel rod active region - 1.6% enr radial 0: Fuel 1.6%" region="-10034" universe="10008" />
<cell id="10054" material="10004" name="Fuel rod active region - 1.6% enr radial 1: Helium" region="10034 -10035" universe="10008" />
<cell id="10055" material="10006" name="Fuel rod active region - 1.6% enr radial 2: Zircaloy 4" region="10035 -10036" universe="10008" />
<cell id="10056" material="10014" name="Fuel rod active region - 1.6% enr radial outer: Borated Water" region="10036" universe="10008" />
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BIN
2x2-periodic/depleted/materials.xml (Stored with Git LFS)
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13
2x2-periodic/depleted/settings.xml
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@ -1,13 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
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11
2x2-periodic/depleted/tallies.xml
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242
2x2-periodic/fresh/geometry.xml
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@ -1,242 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
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<surface coeffs="0.61015" id="10009" name="maximum y" type="y-plane" />
<surface coeffs="-0.61049" id="10010" name="minimum x" type="x-plane" />
<surface coeffs="0.61049" id="10011" name="maximum x" type="x-plane" />
<surface coeffs="-0.61049" id="10012" name="minimum y" type="y-plane" />
<surface coeffs="0.61049" id="10013" name="maximum y" type="y-plane" />
<surface coeffs="37.1621" id="10018" name="Bottom of grid 1" type="z-plane" />
<surface coeffs="40.52" id="10019" name="Top of grid 1" type="z-plane" />
<surface coeffs="98.025" id="10020" name="Bottom of grid 2" type="z-plane" />
<surface coeffs="103.74" id="10021" name="Top of grid 2" type="z-plane" />
<surface coeffs="150.222" id="10022" name="Bottom of grid 3" type="z-plane" />
<surface coeffs="155.937" id="10023" name="Top of grid 3" type="z-plane" />
<surface coeffs="202.419" id="10024" name="Bottom of grid 4" type="z-plane" />
<surface coeffs="208.134" id="10025" name="Top of grid 4" type="z-plane" />
<surface coeffs="254.616" id="10026" name="Bottom of grid 5" type="z-plane" />
<surface coeffs="260.331" id="10027" name="Top of grid 5" type="z-plane" />
<surface coeffs="306.813" id="10028" name="Bottom of grid 6" type="z-plane" />
<surface coeffs="312.528" id="10029" name="Top of grid 6" type="z-plane" />
<surface coeffs="359.01" id="10030" name="Bottom of grid 7" type="z-plane" />
<surface coeffs="364.725" id="10031" name="Top of grid 7" type="z-plane" />
<surface coeffs="411.806" id="10032" name="Bottom of grid 8" type="z-plane" />
<surface coeffs="415.164" id="10033" name="Top of grid 8" type="z-plane" />
<surface coeffs="0.0 0.0 0.39218" id="10034" name="Fuel pellet OR" type="z-cylinder" />
<surface coeffs="0.0 0.0 0.40005" id="10035" name="Fuel clad IR" type="z-cylinder" />
<surface coeffs="0.0 0.0 0.4572" id="10036" name="Fuel clad OR" type="z-cylinder" />
<surface coeffs="0.0 0.0 0.06459" id="10037" name="Fuel rod plenum spring OR" type="z-cylinder" />
<surface coeffs="35.0" id="10038" name="Fuel rod bottom" type="z-plane" />
<surface coeffs="36.748" id="10039" name="Fuel lower fitting top" type="z-plane" />
<surface coeffs="402.508" id="10040" name="Fuel active region top" type="z-plane" />
<surface coeffs="417.164" id="10041" name="Fuel plenum top" type="z-plane" />
<surface coeffs="419.704" id="10042" name="Fuel upper fitting top" type="z-plane" />
<surface coeffs="0.0 0.0 0.56134" id="10043" name="Guide tube IR" type="z-cylinder" />
<surface coeffs="0.0 0.0 0.60198" id="10044" name="Guide tube OR" type="z-cylinder" />
<surface coeffs="0.0 0.0 0.50419" id="10045" name="Guide tube IR below dashpot" type="z-cylinder" />
<surface coeffs="0.0 0.0 0.5461" id="10046" name="Guide tube OR below dashpot" type="z-cylinder" />
<surface coeffs="35.0" id="10047" name="Bottom of GT rod" type="z-plane" />
<surface coeffs="39.958" id="10048" name="GT Dashpot plane" type="z-plane" />
<surface coeffs="423.049" id="10049" name="Top of GT rod" type="z-plane" />
<surface coeffs="0.0 0.0 0.43688" id="10050" name="Instrument tube thimble IR" type="z-cylinder" />
<surface coeffs="0.0 0.0 0.48387" id="10051" name="Instrument tube thimble OR" type="z-cylinder" />
<surface coeffs="0.0 0.0 0.214" id="10052" name="BPRA rod radius 1" type="z-cylinder" />
<surface coeffs="0.0 0.0 0.23051" id="10053" name="BPRA rod radius 2" type="z-cylinder" />
<surface coeffs="0.0 0.0 0.2413" id="10054" name="BPRA rod radius 3" type="z-cylinder" />
<surface coeffs="0.0 0.0 0.42672" id="10055" name="BPRA rod radius 4" type="z-cylinder" />
<surface coeffs="0.0 0.0 0.43688" id="10056" name="BPRA rod radius 5" type="z-cylinder" />
<surface coeffs="0.0 0.0 0.48387" id="10057" name="BPRA rod radius 6" type="z-cylinder" />
<surface coeffs="38.66" id="10058" name="Bottom of BPRA rod" type="z-plane" />
<surface coeffs="40.558" id="10059" name="Top of lower fitting in BPRA rod" type="z-plane" />
<surface coeffs="401.238" id="10060" name="Top of active poison in BPRA rod" type="z-plane" />
<surface coeffs="421.532" id="10061" name="Top of plenum in BPRA rod" type="z-plane" />
<surface coeffs="431.876" id="10062" name="Top of BPRA rod" type="z-plane" />
<surface boundary="periodic" coeffs="-21.41728" id="10149" type="x-plane" />
<surface boundary="periodic" coeffs="21.41728" id="10150" type="x-plane" />
<surface boundary="periodic" coeffs="-21.41728" id="10151" type="y-plane" />
<surface boundary="periodic" coeffs="21.41728" id="10152" type="y-plane" />
<surface boundary="reflective" coeffs="192.5" id="10153" type="z-plane" />
<surface boundary="reflective" coeffs="197.5" id="10154" type="z-plane" />
</geometry>

164
2x2-periodic/fresh/materials.xml
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@ -1,164 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<materials>
<material id="10000" name="Air">
<temperature>300</temperature>
<density units="g/cc" value="0.00616" />
<nuclide ao="0.2094205995" name="O16" />
<nuclide ao="7.94005e-05" name="O17" />
<nuclide ao="0.7780395633" name="N14" />
<nuclide ao="0.0028604367000000003" name="N15" />
<nuclide ao="3.1124879999999996e-05" name="Ar36" />
<nuclide ao="5.86857e-06" name="Ar38" />
<nuclide ao="0.00929300655" name="Ar40" />
<nuclide ao="0.00027" name="C0" />
</material>
<material id="10001" name="SS304">
<temperature>300</temperature>
<density units="g/cc" value="8.03" />
<nuclide name="Si28" wo="0.005512411036974801" />
<nuclide name="Si29" wo="0.0002899050175706448" />
<nuclide name="Si30" wo="0.00019768394545455493" />
<nuclide name="Cr50" wo="0.007930004298910168" />
<nuclide name="Cr52" wo="0.15902878860895933" />
<nuclide name="Cr53" wo="0.018379815074864116" />
<nuclide name="Cr54" wo="0.004661392017266364" />
<nuclide name="Mn55" wo="0.02" />
<nuclide name="Fe54" wo="0.03861561826636726" />
<nuclide name="Fe56" wo="0.6286064568062312" />
<nuclide name="Fe57" wo="0.014776921339264018" />
<nuclide name="Fe58" wo="0.002001003588137652" />
<nuclide name="Ni58" wo="0.06719770531879568" />
<nuclide name="Ni60" wo="0.02677596289274688" />
<nuclide name="Ni61" wo="0.0011833590846680462" />
<nuclide name="Ni62" wo="0.0038348222920813694" />
<nuclide name="Ni64" wo="0.0010081504117080411" />
</material>
<material id="10004" name="Helium">
<temperature>300</temperature>
<density units="g/cc" value="0.0015981" />
<nuclide name="He3" wo="1.5070346049256974e-06" />
<nuclide name="He4" wo="0.999998492965395" />
</material>
<material id="10005" name="Inconel 718">
<temperature>300</temperature>
<density units="g/cc" value="8.2" />
<nuclide name="Si28" wo="0.003215573104901967" />
<nuclide name="Si29" wo="0.00016911126024954278" />
<nuclide name="Si30" wo="0.00011531563484849038" />
<nuclide name="Cr50" wo="0.007913309553017726" />
<nuclide name="Cr52" wo="0.15869399115925625" />
<nuclide name="Cr53" wo="0.018341120727338085" />
<nuclide name="Cr54" wo="0.004651578560387908" />
<nuclide name="Mn55" wo="0.0087" />
<nuclide name="Fe54" wo="0.016163233201258693" />
<nuclide name="Fe56" wo="0.26311407687664323" />
<nuclide name="Fe57" wo="0.006185135350045743" />
<nuclide name="Fe58" wo="0.0008375545720523535" />
<nuclide name="Ni58" wo="0.34398505352691505" />
<nuclide name="Ni60" wo="0.1370661540479713" />
<nuclide name="Ni61" wo="0.006057615154415727" />
<nuclide name="Ni62" wo="0.01963045531316453" />
<nuclide name="Ni64" wo="0.005160721957533462" />
</material>
<material id="10006" name="Zircaloy 4">
<temperature>300</temperature>
<density units="g/cc" value="6.55" />
<nuclide name="O16" wo="0.0012494965182849112" />
<nuclide name="O17" wo="5.034817150887735e-07" />
<nuclide name="Cr50" wo="4.1736864731106146e-05" />
<nuclide name="Cr52" wo="0.0008369936242576807" />
<nuclide name="Cr53" wo="9.673586881507429e-05" />
<nuclide name="Cr54" wo="2.4533642196138756e-05" />
<nuclide name="Fe54" wo="0.00011855672274761877" />
<nuclide name="Fe56" wo="0.001929932104229657" />
<nuclide name="Fe57" wo="4.536774095388075e-05" />
<nuclide name="Fe58" wo="6.143432068843669e-06" />
<nuclide name="Zr90" wo="0.49750307249921255" />
<nuclide name="Zr91" wo="0.10970127796055709" />
<nuclide name="Zr92" wo="0.16952409354767467" />
<nuclide name="Zr94" wo="0.17553856942304608" />
<nuclide name="Zr96" wo="0.02888298656950975" />
<nuclide name="Sn112" wo="0.0001325869644430062" />
<nuclide name="Sn114" wo="9.182449637587617e-05" />
<nuclide name="Sn115" wo="4.771905922545867e-05" />
<nuclide name="Sn116" wo="0.002058423153629443" />
<nuclide name="Sn117" wo="0.0010966473429083066" />
<nuclide name="Sn118" wo="0.0034879812938438245" />
<nuclide name="Sn119" wo="0.001247577110245757" />
<nuclide name="Sn120" wo="0.004771539495238715" />
<nuclide name="Sn122" wo="0.0006894094798456136" />
<nuclide name="Sn124" wo="0.000876291604244001" />
</material>
<material id="10008" name="Fuel 1.6%">
<temperature>300</temperature>
<density units="g/cc" value="10.31341" />
<nuclide ao="1.9992419999999993" name="O16" />
<nuclide ao="0.0007579999999999998" name="O17" />
<nuclide ao="0.00013098435147670763" name="U234" />
<nuclide ao="0.01630317699531038" name="U235" />
<nuclide ao="0.9835658386532129" name="U238" />
</material>
<material id="10010" name="Fuel 3.1%">
<temperature>300</temperature>
<density units="g/cc" value="10.30166" />
<nuclide ao="1.9992420000000022" name="O16" />
<nuclide ao="0.0007580000000000009" name="O17" />
<nuclide ao="0.0002523276152188021" name="U234" />
<nuclide ao="0.03140636057161555" name="U235" />
<nuclide ao="0.9683413118131656" name="U238" />
</material>
<material id="10013" name="Borosilicate Glass">
<temperature>300</temperature>
<density units="g/cc" value="2.26" />
<nuclide ao="0.013479369482225239" name="B10" />
<nuclide ao="0.054735873774104264" name="B11" />
<nuclide ao="0.6509787013744828" name="O16" />
<nuclide ao="0.00024681447050525047" name="O17" />
<nuclide ao="0.23640592474731761" name="Si28" />
<nuclide ao="0.01200401834354893" name="Si29" />
<nuclide ao="0.007913102535354443" name="Si30" />
<nuclide ao="0.024236195272461444" name="Al27" />
</material>
<material id="10014" name="Borated Water">
<temperature>300</temperature>
<density units="g/cc" value="0.7405820675158279" />
<nuclide ao="0.00032178659941803253" name="B10" />
<nuclide ao="0.0013017583017829388" name="B11" />
<nuclide ao="1.996441935899364" name="H1" />
<nuclide ao="0.0003109742982341739" name="H2" />
<nuclide ao="0.9979980704223166" name="O16" />
<nuclide ao="0.0003783846764824448" name="O17" />
<sab name="c_H_in_H2O" />
</material>
<material id="10015" name="Water SPN">
<density units="g/cc" value="0.9810025319057221" />
<nuclide ao="0.00032178659941803253" name="B10" />
<nuclide ao="0.0013017583017829388" name="B11" />
<nuclide ao="1.996441935899364" name="H1" />
<nuclide ao="0.0003109742982341739" name="H2" />
<nuclide ao="0.9979980704223166" name="O16" />
<nuclide ao="0.0003783846764824448" name="O17" />
<sab name="c_H_in_H2O" />
</material>
<material id="10016" name="SS SPN">
<temperature>300</temperature>
<density units="g/cc" value="3.6838480704877297" />
<nuclide name="Si28" wo="0.005512411036974801" />
<nuclide name="Si29" wo="0.0002899050175706448" />
<nuclide name="Si30" wo="0.00019768394545455493" />
<nuclide name="Cr50" wo="0.007930004298910168" />
<nuclide name="Cr52" wo="0.15902878860895933" />
<nuclide name="Cr53" wo="0.018379815074864116" />
<nuclide name="Cr54" wo="0.004661392017266364" />
<nuclide name="Mn55" wo="0.02" />
<nuclide name="Fe54" wo="0.03861561826636726" />
<nuclide name="Fe56" wo="0.6286064568062312" />
<nuclide name="Fe57" wo="0.014776921339264018" />
<nuclide name="Fe58" wo="0.002001003588137652" />
<nuclide name="Ni58" wo="0.06719770531879568" />
<nuclide name="Ni60" wo="0.02677596289274688" />
<nuclide name="Ni61" wo="0.0011833590846680462" />
<nuclide name="Ni62" wo="0.0038348222920813694" />
<nuclide name="Ni64" wo="0.0010081504117080411" />
</material>
</materials>

8
2x2-periodic/fresh/plots.xml
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@ -1,8 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<plots>
<plot basis="xy" color_by="material" filename="2x2-periodic" id="10020" type="slice">
<origin>0.0 0.0 195.0</origin>
<width>42.83456 42.83456</width>
<pixels>1000 1000</pixels>
</plot>
</plots>

15
2x2-periodic/fresh/settings.xml
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@ -1,15 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<settings>
<run_mode>eigenvalue</run_mode>
<particles>10000</particles>
<batches>10</batches>
<inactive>5</inactive>
<source strength="1.0">
<space type="fission">
<parameters>-21.41728 -21.41728 192.5 21.41728 21.41728 197.5</parameters>
</space>
</source>
<output>
<tallies>false</tallies>
</output>
</settings>

19
2x2-periodic/fresh/tallies.xml
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@ -1,19 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<tallies>
<filter id="1" type="distribcell">
<bins>10053</bins>
</filter>
<filter id="2" type="distribcell">
<bins>10061</bins>
</filter>
<tally id="10000" name="depletion tally">
<filters>1</filters>
<nuclides>O16 O17 U234 U235 U238</nuclides>
<scores>(n,p) (n,a) (n,gamma) fission (n,2n) (n,3n) (n,4n)</scores>
</tally>
<tally id="10001" name="depletion tally">
<filters>2</filters>
<nuclides>O16 O17 U234 U235 U238</nuclides>
<scores>(n,p) (n,a) (n,gamma) fission (n,2n) (n,3n) (n,4n)</scores>
</tally>
</tallies>

142
2x2-periodic/geometry.py
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@ -1,142 +0,0 @@
import openmc
from beavrs.builder import BEAVRS
def find_assembly(assembly_name, wrap_geometry=True):
"""Find a fuel assembly with some string name in the BEAVRS OpenMC model.
This method extracts the fuel assembly and wraps it in an OpenMC Geometry.
The returned geometry has reflective boundary conditions along all
boundaries. The z-axis is bounded between z=200 and z=210 cm.
Parameters
----------
assembly_name : str
The name of the fuel assembly lattice
wrap_geometry : bool
If false, the fuel assembly Lattice is returned. If true, the fuel
assembly Lattice is wrapped in an OpenMC Geometry and returned (default).
Returns
-------
fuel_assembly
The OpenMC Lattice or Geometry for the assembly or None if not found
"""
# Get OpenMC Lattices for the fuel assembly
fuel_assembly = \
beavrs.openmc_geometry.get_lattices_by_name(assembly_name)[0]
# Wrap lattice in a Geometry if requested by the user
if wrap_geometry:
# Create a root Cell
root_cell = openmc.Cell(name='root cell')
root_cell.fill = fuel_assembly
# Make mixed reflective / vacuum boundaries
min_x = openmc.XPlane(x0=-10.70864, boundary_type='reflective')
max_x = openmc.XPlane(x0=+10.70864, boundary_type='reflective')
min_y = openmc.YPlane(y0=-10.70864, boundary_type='reflective')
max_y = openmc.YPlane(y0=+10.70864, boundary_type='reflective')
max_z = openmc.ZPlane(z0=197.5, boundary_type='reflective')
min_z = openmc.ZPlane(z0=192.5, boundary_type='reflective')
# Add boundaries to the root Cell
root_cell.region = \
+min_x & -max_x & +min_y & -max_y & +min_z & -max_z
# Create a root Universe
root_univ = openmc.Universe(universe_id=0, name='root universe')
root_univ.add_cell(root_cell)
# Create a Geometry
fuel_assembly = openmc.Geometry(root_univ)
return fuel_assembly
def build_two_by_two(assembly1_name, assembly2_name):
"""Build a 2x2 fuel assembly geometry.
This routine puts reflective boundary conditions along all boundaries.
Parameters
----------
assembly1_name : str
The BEAVRS fuel assembly to place in the bottom right and top left
assembly2_name : str
The BEAVRS fuel assembly to place in the bottom left and top right
Returns
-------
openmc.Geometry
A 2x2 fuel assembly OpenMC Geometry
"""
fuel_assembly1 = find_assembly(assembly1_name, wrap_geometry=False)
fuel_assembly2 = find_assembly(assembly2_name, wrap_geometry=False)
# Find the water material
all_cells = beavrs.main_universe.get_all_cells()
for cell_uuid, cell in all_cells.items():
if cell.fill_type == 'material' and cell.fill.name == 'Borated Water':
water = cell.fill
# Create a Cell/Universe around the first fuel assembly
fuel_cell1 = openmc.Cell(name='assm1 cell')
fuel_cell1.fill = fuel_assembly1
fuel_univ1 = openmc.Universe(name='assm1 universe')
fuel_univ1.add_cell(fuel_cell1)
# Create a Cell/Universe around the second fuel assembly
fuel_cell2 = openmc.Cell(name='assm2 cell')
fuel_cell2.fill = fuel_assembly2
fuel_univ2 = openmc.Universe(name='assm2 universe')
fuel_univ2.add_cell(fuel_cell2)
# Create a 3x3 lattice two fuel assemblies surrounded by a water reflector
two_by_two_lattice = openmc.RectLattice(name='reflector')
two_by_two_lattice.lower_left = [-21.41728, -21.41728, -500.]
two_by_two_lattice.pitch = [21.41728, 21.41728, 1000.]
two_by_two_lattice.universes = [[[fuel_univ1, fuel_univ2],
[fuel_univ2, fuel_univ1]]]
# Create a Geometry around the reflected lattice
root_cell = openmc.Cell(name='root cell')
root_cell.fill = two_by_two_lattice
# Make mixed reflective / vacuum boundaries
min_x = openmc.XPlane(x0=-21.41728, boundary_type='periodic')
max_x = openmc.XPlane(x0=+21.41728, boundary_type='periodic')
min_y = openmc.YPlane(y0=-21.41728, boundary_type='periodic')
max_y = openmc.YPlane(y0=+21.41728, boundary_type='periodic')
min_z = openmc.ZPlane(z0=192.5, boundary_type='reflective')
max_z = openmc.ZPlane(z0=197.5, boundary_type='reflective')
# Add boundaries to the root Cell
root_cell.region = +min_x & -max_x & +min_y & -max_y & +min_z & -max_z
# Create a root Universe for this fuel assembly
root_univ = openmc.Universe(universe_id=0, name='root universe')
root_univ.add_cell(root_cell)
# Create an OpenMC Geometry for this fuel assembly
two_by_two = openmc.Geometry(root_univ)
return two_by_two
#### Create OpenMC "materials.xml" and "geometry.xml" files
# Instantiate a BEAVRS object
beavrs = BEAVRS()
# Write all BEAVRS materials to materials.xml file
beavrs.write_openmc_materials()
# Extract fuel assemblies of interest from BEAVRS model
openmc_geometry = build_two_by_two('Fuel 1.6% enr instr no BAs',
'Fuel 3.1% enr instr 20')

59
2x2-reflector/build-depleted.py
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@ -1,59 +0,0 @@
#!/usr/bin/env python3
import numpy as np
import opendeplete
from geometry import openmc_geometry
# FIXME: Automatically extract info needed to calculate burnable cell volumes
# Fuel rod geometric parameters
radius = 0.39218
height = 5.
# Count the number of instances for each cell and material
openmc_geometry.determine_paths(instances_only=True)
# Extract all cells filled by a fuel material
fuel_cells = openmc_geometry.get_cells_by_name(
name='enr radial 0: Fuel', case_sensitive=True)
# Assign distribmats for each material
for cell in fuel_cells:
cell.fill.volume = np.pi * radius**2 * height
cell.fill.depletable = True
cell.fill.temperature = 300.0
cell.fill = [cell.fill.clone() for i in range(cell.num_instances)]
# Set temperature for all cells
cells = openmc_geometry.get_all_cells()
for cell_id in cells:
cells[cell_id].temperature = 300.0
# Create dt vector for 1 month with 5 day timesteps
dt1 = 5*24*60*60 # 5 days
dt2 = 1.*30*24*60*60 # 1 months
N = np.floor(dt2/dt1)
dt = np.repeat([dt1], N)
# Create settings variable
settings = opendeplete.OpenMCSettings()
settings.openmc_call = "openmc"
settings.particles = 120000
settings.batches = 30
settings.inactive = 20
settings.lower_left = [-32.12592, -32.12592, 192.5]
settings.upper_right = [32.12592, 32.12592, 197.5]
settings.entropy_dimension = [17*3, 17*3, 1]
# MeV/second cm from CASMO
settings.power = 2.337e15 * ((17.*17.*2.) / 1.5**2) * height
settings.dt_vec = dt
settings.output_dir = 'depleted'
op = opendeplete.OpenMCOperator(openmc_geometry, settings)
# Perform simulation using the MCNPX/MCNP6 algorithm
opendeplete.cecm(op)

143
2x2-reflector/build-fresh.py
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@ -1,143 +0,0 @@
#!/usr/bin/env python3
"""Creates a 2D 2x2 assembly colorset with a water reflector."""
import os
import shutil
import copy
import numpy as np
import openmc
from geometry import beavrs, openmc_geometry
#### Query the user for options
# Query the user on whether to use multipole cross sections
multipole = input('Use multipole cross sections? (y/n): ').lower()
multipole = (multipole == 'y')
# Query the user on whether to use distribmats or distribcells
# If using distribmats, the geometry must be "differentiated" with unique
# material instances for each instance of a fuel cell
distrib = input('Use distribmat or distribcells? [mat/cell]: ').lower()
if distrib not in ['cell', 'mat']:
raise InputError('Distrib type "{}" is unsupported'.format(distrib))
#### "Differentiate" the geometry if using distribmats
if distrib == 'mat':
# Count the number of instances for each cell and material
openmc_geometry.determine_paths()
# Extract all cells filled by a fuel material
fuel_cells = openmc_geometry.get_cells_by_name(
name='enr radial 0: Fuel', case_sensitive=True)
# Assign distribmats for each material
for cell in fuel_cells:
new_materials = []
for i in range(cell.num_instances):
new_materials.append(cell.fill.clone())
# Fill cell with list of "differentiated" materials
cell.fill = new_materials
#### Create OpenMC "materials.xml" file
all_materials = openmc_geometry.get_all_materials()
materials = openmc.Materials(all_materials.values())
materials.export_to_xml()
#### Create OpenMC "geometry.xml" file
openmc_geometry.export_to_xml()
#### Create OpenMC "settings.xml" file
# Construct uniform initial source distribution over fissionable zones
lower_left = [-32.12592, -32.12592, 192.5]
upper_right = [32.12592, 32.12592, 197.5]
lat_width = (np.array(upper_right) - np.array(lower_left))
lat_width[:2] /= 3.
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
source.space.only_fissionable = True
settings_file = openmc.Settings()
settings_file.batches = 10
settings_file.inactive = 5
settings_file.particles = 10000
settings_file.output = {'tallies': False}
settings_file.source = source
settings_file.sourcepoint_write = False
if multipole:
settings_file.temperature = {'multipole': True, 'tolerance': 1000}
settings_file.export_to_xml()
#### Create OpenMC "plots.xml" file
# Initialize the BEAVRS color mapping scheme
beavrs.write_openmc_plots()
# Create a plot colored by materials
plot = openmc.Plot()
plot.width = [64.25184, 64.25184]
plot.origin = [0., 0., 195.]
plot.color_by = 'material'
plot.filename = '2x2-reflector'
plot.pixels = [1000, 1000]
plot_file = openmc.Plots([plot])
plot_file.export_to_xml()
#### Create OpenMC "tallies.xml" file
tallies = openmc.Tallies()
# Extract all fuel materials
materials = openmc_geometry.get_materials_by_name(name='Fuel', matching=False)
# If using distribcells, create distribcell tally needed for depletion
if distrib == 'cell':
fuel_cells = openmc_geometry.get_cells_by_name(
name='enr radial 0: Fuel', case_sensitive=True, matching=False)
for cell in fuel_cells:
tally = openmc.Tally(name='depletion tally')
tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
'fission', '(n,2n)', '(n,3n)', '(n,4n)']
tally.nuclides = cell.fill.get_nuclides()
tally.filters.append(openmc.DistribcellFilter([cell.id]))
tallies.append(tally)
# If using distribmats, create material tally needed for depletion
elif distrib == 'mat':
tally = openmc.Tally(name='depletion tally')
tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
'fission', '(n,2n)', '(n,3n)', '(n,4n)']
tally.nuclides = materials[0].get_nuclides()
material_ids = [material.id for material in materials]
tally.filters.append(openmc.MaterialFilter(material_ids))
tallies.append(tally)
tallies.export_to_xml()
#### Move all XML files to 'fresh' directory
if not os.path.exists('fresh'):
os.makedirs('fresh')
shutil.move('materials.xml', 'fresh/materials.xml')
shutil.move('geometry.xml', 'fresh/geometry.xml')
shutil.move('settings.xml', 'fresh/settings.xml')
shutil.move('tallies.xml', 'fresh/tallies.xml')
shutil.move('plots.xml', 'fresh/plots.xml')

244
2x2-reflector/depleted/geometry.xml
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@ -1,244 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="10001" material="10014" name="Intermediate grid pincell radial 0: Borated Water" region="10010 -10011 10012 -10013" universe="10001" />
<cell id="10002" material="10006" name="Intermediate grid pincell radial outer: Zircaloy 4" region="~(10010 -10011 10012 -10013)" universe="10001" />
<cell id="10003" material="10014" name="Top/Bottom grid pincell radial 0: Borated Water" region="10006 -10007 10008 -10009" universe="10002" />
<cell id="10004" material="10005" name="Top/Bottom grid pincell radial outer: Inconel 718" region="~(10006 -10007 10008 -10009)" universe="10002" />
<cell id="10009" material="10014" name="Grids axial universe axial 0: Borated Water" region="-10000" universe="10005" />
<cell id="10010" material="10015" name="Grids axial universe axial 1: Water SPN" region="-10003 10000" universe="10005" />
<cell id="10011" material="10014" name="Grids axial universe axial 2: Borated Water" region="-10018 10003" universe="10005" />
<cell fill="10002" id="10012" name="Grids axial universe axial 3: Top/Bottom grid pincell" region="-10019 10018" universe="10005" />
<cell id="10013" material="10014" name="Grids axial universe axial 4: Borated Water" region="-10020 10019" universe="10005" />
<cell fill="10001" id="10014" name="Grids axial universe axial 5: Intermediate grid pincell" region="-10021 10020" universe="10005" />
<cell id="10015" material="10014" name="Grids axial universe axial 6: Borated Water" region="-10022 10021" universe="10005" />
<cell fill="10001" id="10016" name="Grids axial universe axial 7: Intermediate grid pincell" region="-10023 10022" universe="10005" />
<cell id="10017" material="10014" name="Grids axial universe axial 8: Borated Water" region="-10024 10023" universe="10005" />
<cell fill="10001" id="10018" name="Grids axial universe axial 9: Intermediate grid pincell" region="-10025 10024" universe="10005" />
<cell id="10019" material="10014" name="Grids axial universe axial 10: Borated Water" region="-10026 10025" universe="10005" />
<cell fill="10001" id="10020" name="Grids axial universe axial 11: Intermediate grid pincell" region="-10027 10026" universe="10005" />
<cell id="10021" material="10014" name="Grids axial universe axial 12: Borated Water" region="-10028 10027" universe="10005" />
<cell fill="10001" id="10022" name="Grids axial universe axial 13: Intermediate grid pincell" region="-10029 10028" universe="10005" />
<cell id="10023" material="10014" name="Grids axial universe axial 14: Borated Water" region="-10030 10029" universe="10005" />
<cell fill="10001" id="10024" name="Grids axial universe axial 15: Intermediate grid pincell" region="-10031 10030" universe="10005" />
<cell id="10025" material="10014" name="Grids axial universe axial 16: Borated Water" region="-10032 10031" universe="10005" />
<cell fill="10002" id="10026" name="Grids axial universe axial 17: Top/Bottom grid pincell" region="-10033 10032" universe="10005" />
<cell id="10027" material="10014" name="Grids axial universe axial 18: Borated Water" region="-10004 10033" universe="10005" />
<cell id="10028" material="10015" name="Grids axial universe axial 19: Water SPN" region="-10005 10004" universe="10005" />
<cell id="10029" material="10014" name="Grids axial universe axial top: Borated Water" region="10005" universe="10005" />
<cell id="10051" material="10006" name="Fuel rod lower/upper fitting radial 0: Zircaloy 4" region="-10036" universe="10007" />
<cell id="10052" material="10014" name="Fuel rod lower/upper fitting radial outer: Borated Water" region="10036" universe="10007" />
<cell id="10053" material="10017 10018 10019 10020 10021 10022 10023 10024 10025 10026 10027 10028 10029 10030 10031 10032 10033 10034 10035 10036 10037 10038 10039 10040 10041 10042 10043 10044 10045 10046 10047 10048 10049 10050 10051 10052 10053 10054 10055 10056 10057 10058 10059 10060 10061 10062 10063 10064 10065 10066 10067 10068 10069 10070 10071 10072 10073 10074 10075 10076 10077 10078 10079 10080 10081 10082 10083 10084 10085 10086 10087 10088 10089 10090 10091 10092 10093 10094 10095 10096 10097 10098 10099 10100 10101 10102 10103 10104 10105 10106 10107 10108 10109 10110 10111 10112 10113 10114 10115 10116 10117 10118 10119 10120 10121 10122 10123 10124 10125 10126 10127 10128 10129 10130 10131 10132 10133 10134 10135 10136 10137 10138 10139 10140 10141 10142 10143 10144 10145 10146 10147 10148 10149 10150 10151 10152 10153 10154 10155 10156 10157 10158 10159 10160 10161 10162 10163 10164 10165 10166 10167 10168 10169 10170 10171 10172 10173 10174 10175 10176 10177 10178 10179 10180 10181 10182 10183 10184 10185 10186 10187 10188 10189 10190 10191 10192 10193 10194 10195 10196 10197 10198 10199 10200 10201 10202 10203 10204 10205 10206 10207 10208 10209 10210 10211 10212 10213 10214 10215 10216 10217 10218 10219 10220 10221 10222 10223 10224 10225 10226 10227 10228 10229 10230 10231 10232 10233 10234 10235 10236 10237 10238 10239 10240 10241 10242 10243 10244 10245 10246 10247 10248 10249 10250 10251 10252 10253 10254 10255 10256 10257 10258 10259 10260 10261 10262 10263 10264 10265 10266 10267 10268 10269 10270 10271 10272 10273 10274 10275 10276 10277 10278 10279 10280 10281 10282 10283 10284 10285 10286 10287 10288 10289 10290 10291 10292 10293 10294 10295 10296 10297 10298 10299 10300 10301 10302 10303 10304 10305 10306 10307 10308 10309 10310 10311 10312 10313 10314 10315 10316 10317 10318 10319 10320 10321 10322 10323 10324 10325 10326 10327 10328 10329 10330 10331 10332 10333 10334 10335 10336 10337 10338 10339 10340 10341 10342 10343 10344 10345 10346 10347 10348 10349 10350 10351 10352 10353 10354 10355 10356 10357 10358 10359 10360 10361 10362 10363 10364 10365 10366 10367 10368 10369 10370 10371 10372 10373 10374 10375 10376 10377 10378 10379 10380 10381 10382 10383 10384 10385 10386 10387 10388 10389 10390 10391 10392 10393 10394 10395 10396 10397 10398 10399 10400 10401 10402 10403 10404 10405 10406 10407 10408 10409 10410 10411 10412 10413 10414 10415 10416 10417 10418 10419 10420 10421 10422 10423 10424 10425 10426 10427 10428 10429 10430 10431 10432 10433 10434 10435 10436 10437 10438 10439 10440 10441 10442 10443 10444 10445 10446 10447 10448 10449 10450 10451 10452 10453 10454 10455 10456 10457 10458 10459 10460 10461 10462 10463 10464 10465 10466 10467 10468 10469 10470 10471 10472 10473 10474 10475 10476 10477 10478 10479 10480 10481 10482 10483 10484 10485 10486 10487 10488 10489 10490 10491 10492 10493 10494 10495 10496 10497 10498 10499 10500 10501 10502 10503 10504 10505 10506 10507 10508 10509 10510 10511 10512 10513 10514 10515 10516 10517 10518 10519 10520 10521 10522 10523 10524 10525 10526 10527 10528 10529 10530 10531 10532 10533 10534 10535 10536 10537 10538 10539 10540 10541 10542 10543 10544" name="Fuel rod active region - 1.6% enr radial 0: Fuel 1.6%" region="-10034" universe="10008" />
<cell id="10054" material="10004" name="Fuel rod active region - 1.6% enr radial 1: Helium" region="10034 -10035" universe="10008" />
<cell id="10055" material="10006" name="Fuel rod active region - 1.6% enr radial 2: Zircaloy 4" region="10035 -10036" universe="10008" />
<cell id="10056" material="10014" name="Fuel rod active region - 1.6% enr radial outer: Borated Water" region="10036" universe="10008" />
<cell id="10061" material="10545 10546 10547 10548 10549 10550 10551 10552 10553 10554 10555 10556 10557 10558 10559 10560 10561 10562 10563 10564 10565 10566 10567 10568 10569 10570 10571 10572 10573 10574 10575 10576 10577 10578 10579 10580 10581 10582 10583 10584 10585 10586 10587 10588 10589 10590 10591 10592 10593 10594 10595 10596 10597 10598 10599 10600 10601 10602 10603 10604 10605 10606 10607 10608 10609 10610 10611 10612 10613 10614 10615 10616 10617 10618 10619 10620 10621 10622 10623 10624 10625 10626 10627 10628 10629 10630 10631 10632 10633 10634 10635 10636 10637 10638 10639 10640 10641 10642 10643 10644 10645 10646 10647 10648 10649 10650 10651 10652 10653 10654 10655 10656 10657 10658 10659 10660 10661 10662 10663 10664 10665 10666 10667 10668 10669 10670 10671 10672 10673 10674 10675 10676 10677 10678 10679 10680 10681 10682 10683 10684 10685 10686 10687 10688 10689 10690 10691 10692 10693 10694 10695 10696 10697 10698 10699 10700 10701 10702 10703 10704 10705 10706 10707 10708 10709 10710 10711 10712 10713 10714 10715 10716 10717 10718 10719 10720 10721 10722 10723 10724 10725 10726 10727 10728 10729 10730 10731 10732 10733 10734 10735 10736 10737 10738 10739 10740 10741 10742 10743 10744 10745 10746 10747 10748 10749 10750 10751 10752 10753 10754 10755 10756 10757 10758 10759 10760 10761 10762 10763 10764 10765 10766 10767 10768 10769 10770 10771 10772 10773 10774 10775 10776 10777 10778 10779 10780 10781 10782 10783 10784 10785 10786 10787 10788 10789 10790 10791 10792 10793 10794 10795 10796 10797 10798 10799 10800 10801 10802 10803 10804 10805 10806 10807 10808 10809 10810 10811 10812 10813 10814 10815 10816 10817 10818 10819 10820 10821 10822 10823 10824 10825 10826 10827 10828 10829 10830 10831 10832 10833 10834 10835 10836 10837 10838 10839 10840 10841 10842 10843 10844 10845 10846 10847 10848 10849 10850 10851 10852 10853 10854 10855 10856 10857 10858 10859 10860 10861 10862 10863 10864 10865 10866 10867 10868 10869 10870 10871 10872 10873 10874 10875 10876 10877 10878 10879 10880 10881 10882 10883 10884 10885 10886 10887 10888 10889 10890 10891 10892 10893 10894 10895 10896 10897 10898 10899 10900 10901 10902 10903 10904 10905 10906 10907 10908 10909 10910 10911 10912 10913 10914 10915 10916 10917 10918 10919 10920 10921 10922 10923 10924 10925 10926 10927 10928 10929 10930 10931 10932 10933 10934 10935 10936 10937 10938 10939 10940 10941 10942 10943 10944 10945 10946 10947 10948 10949 10950 10951 10952 10953 10954 10955 10956 10957 10958 10959 10960 10961 10962 10963 10964 10965 10966 10967 10968 10969 10970 10971 10972 10973 10974 10975 10976 10977 10978 10979 10980 10981 10982 10983 10984 10985 10986 10987 10988 10989 10990 10991 10992 10993 10994 10995 10996 10997 10998 10999 11000 11001 11002 11003 11004 11005 11006 11007 11008 11009 11010 11011 11012 11013 11014 11015 11016 11017 11018 11019 11020 11021 11022 11023 11024 11025 11026 11027 11028 11029 11030 11031 11032 11033 11034 11035 11036 11037 11038 11039 11040 11041 11042 11043 11044 11045 11046 11047 11048 11049 11050 11051 11052 11053 11054 11055 11056 11057 11058 11059 11060 11061 11062 11063 11064 11065 11066 11067 11068 11069 11070 11071 11072" name="Fuel rod active region - 3.1% enr radial 0: Fuel 3.1%" region="-10034" universe="10010" />
<cell id="10062" material="10004" name="Fuel rod active region - 3.1% enr radial 1: Helium" region="10034 -10035" universe="10010" />
<cell id="10063" material="10006" name="Fuel rod active region - 3.1% enr radial 2: Zircaloy 4" region="10035 -10036" universe="10010" />
<cell id="10064" material="10014" name="Fuel rod active region - 3.1% enr radial outer: Borated Water" region="10036" universe="10010" />
<cell id="10073" material="10005" name="Fuel rod plenum radial 0: Inconel 718" region="-10037" universe="10013" />
<cell id="10074" material="10004" name="Fuel rod plenum radial 1: Helium" region="10037 -10035" universe="10013" />
<cell id="10075" material="10006" name="Fuel rod plenum radial 2: Zircaloy 4" region="10035 -10036" universe="10013" />
<cell id="10076" material="10014" name="Fuel rod plenum radial outer: Borated Water" region="10036" universe="10013" />
<cell id="10077" material="10014" name="Fuel rod - 1.6% enr axial 0: Borated Water" region="-10000" universe="10014" />
<cell id="10078" material="10016" name="Fuel rod - 1.6% enr axial 1: SS SPN" region="-10038 10000" universe="10014" />
<cell fill="10007" id="10079" name="Fuel rod - 1.6% enr axial 2: Fuel rod lower/upper fitting" region="-10039 10038" universe="10014" />
<cell fill="10008" id="10080" name="Fuel rod - 1.6% enr axial 3: Fuel rod active region - 1.6% enr" region="-10040 10039" universe="10014" />
<cell fill="10013" id="10081" name="Fuel rod - 1.6% enr axial 4: Fuel rod plenum" region="-10041 10040" universe="10014" />
<cell fill="10007" id="10082" name="Fuel rod - 1.6% enr axial 5: Fuel rod lower/upper fitting" region="-10042 10041" universe="10014" />
<cell id="10083" material="10014" name="Fuel rod - 1.6% enr axial 6: Borated Water" region="-10004 10042" universe="10014" />
<cell id="10084" material="10016" name="Fuel rod - 1.6% enr axial 7: SS SPN" region="-10005 10004" universe="10014" />
<cell id="10085" material="10014" name="Fuel rod - 1.6% enr axial top: Borated Water" region="10005" universe="10014" />
<cell fill="10014" id="10086" name="(Fuel rod - 1.6% enr) wrapped by (Grids axial universe) radial 0: Fuel rod - 1.6% enr" region="-10036" universe="10015" />
<cell fill="10005" id="10087" name="(Fuel rod - 1.6% enr) wrapped by (Grids axial universe) radial outer: Grids axial universe" region="10036" universe="10015" />
<cell id="10099" material="10014" name="Fuel rod - 3.1% enr axial 0: Borated Water" region="-10000" universe="10018" />
<cell id="10100" material="10016" name="Fuel rod - 3.1% enr axial 1: SS SPN" region="-10038 10000" universe="10018" />
<cell fill="10007" id="10101" name="Fuel rod - 3.1% enr axial 2: Fuel rod lower/upper fitting" region="-10039 10038" universe="10018" />
<cell fill="10010" id="10102" name="Fuel rod - 3.1% enr axial 3: Fuel rod active region - 3.1% enr" region="-10040 10039" universe="10018" />
<cell fill="10013" id="10103" name="Fuel rod - 3.1% enr axial 4: Fuel rod plenum" region="-10041 10040" universe="10018" />
<cell fill="10007" id="10104" name="Fuel rod - 3.1% enr axial 5: Fuel rod lower/upper fitting" region="-10042 10041" universe="10018" />
<cell id="10105" material="10014" name="Fuel rod - 3.1% enr axial 6: Borated Water" region="-10004 10042" universe="10018" />
<cell id="10106" material="10016" name="Fuel rod - 3.1% enr axial 7: SS SPN" region="-10005 10004" universe="10018" />
<cell id="10107" material="10014" name="Fuel rod - 3.1% enr axial top: Borated Water" region="10005" universe="10018" />
<cell fill="10018" id="10108" name="(Fuel rod - 3.1% enr) wrapped by (Grids axial universe) radial 0: Fuel rod - 3.1% enr" region="-10036" universe="10019" />
<cell fill="10005" id="10109" name="(Fuel rod - 3.1% enr) wrapped by (Grids axial universe) radial outer: Grids axial universe" region="10036" universe="10019" />
<cell id="10132" material="10014" name="Empty GT below the dashpot radial 0: Borated Water" region="-10045" universe="10024" />
<cell id="10133" material="10006" name="Empty GT below the dashpot radial 1: Zircaloy 4" region="10045 -10046" universe="10024" />
<cell id="10134" material="10014" name="Empty GT below the dashpot radial outer: Borated Water" region="10046" universe="10024" />
<cell id="10135" material="10014" name="Empty GT above the dashpot radial 0: Borated Water" region="-10043" universe="10025" />
<cell id="10136" material="10006" name="Empty GT above the dashpot radial 1: Zircaloy 4" region="10043 -10044" universe="10025" />
<cell id="10137" material="10014" name="Empty GT above the dashpot radial outer: Borated Water" region="10044" universe="10025" />
<cell id="10138" material="10014" name="Empty Guide Tube axial 0: Borated Water" region="-10000" universe="10026" />
<cell id="10139" material="10015" name="Empty Guide Tube axial 1: Water SPN" region="-10047 10000" universe="10026" />
<cell fill="10024" id="10140" name="Empty Guide Tube axial 2: Empty GT below the dashpot" region="-10048 10047" universe="10026" />
<cell fill="10025" id="10141" name="Empty Guide Tube axial 3: Empty GT above the dashpot" region="-10049 10048" universe="10026" />
<cell id="10142" material="10015" name="Empty Guide Tube axial 4: Water SPN" region="-10005 10049" universe="10026" />
<cell id="10143" material="10014" name="Empty Guide Tube axial top: Borated Water" region="10005" universe="10026" />
<cell fill="10026" id="10144" name="(Empty Guide Tube) wrapped by (Grids axial universe) radial 0: Empty Guide Tube" region="-10044" universe="10027" />
<cell fill="10005" id="10145" name="(Empty Guide Tube) wrapped by (Grids axial universe) radial outer: Grids axial universe" region="10044" universe="10027" />
<cell id="10146" material="10014" name="Empty Guide Tube in Center Position axial 0: Borated Water" region="-10000" universe="10028" />
<cell id="10147" material="10015" name="Empty Guide Tube in Center Position axial 1: Water SPN" region="-10047 10000" universe="10028" />
<cell fill="10025" id="10148" name="Empty Guide Tube in Center Position axial 2: Empty GT above the dashpot" region="-10049 10047" universe="10028" />
<cell id="10149" material="10015" name="Empty Guide Tube in Center Position axial 3: Water SPN" region="-10005 10049" universe="10028" />
<cell id="10150" material="10014" name="Empty Guide Tube in Center Position axial top: Borated Water" region="10005" universe="10028" />
<cell fill="10028" id="10151" name="(Empty Guide Tube in Center Position) wrapped by (Grids axial universe) radial 0: Empty Guide Tube in Center Position" region="-10044" universe="10029" />
<cell fill="10005" id="10152" name="(Empty Guide Tube in Center Position) wrapped by (Grids axial universe) radial outer: Grids axial universe" region="10044" universe="10029" />
<cell id="10153" material="10000" name="Instrument tube thimble radial 0: Air" region="-10050" universe="10030" />
<cell id="10154" material="10006" name="Instrument tube thimble radial 1: Zircaloy 4" region="10050 -10051" universe="10030" />
<cell id="10155" material="10014" name="Instrument tube thimble radial outer: Borated Water" region="10051" universe="10030" />
<cell id="10156" material="10000" name="Instrument tube thimble support plane radial 0: Air" region="-10050" universe="10031" />
<cell id="10157" material="10006" name="Instrument tube thimble support plane radial 1: Zircaloy 4" region="10050 -10051" universe="10031" />
<cell id="10158" material="10015" name="Instrument tube thimble support plane radial outer: Water SPN" region="10051" universe="10031" />
<cell fill="10030" id="10159" name="Instrument tube axial stack axial 0: Instrument tube thimble" region="-10000" universe="10032" />
<cell fill="10031" id="10160" name="Instrument tube axial stack axial 1: Instrument tube thimble support plane" region="-10038 10000" universe="10032" />
<cell fill="10030" id="10161" name="Instrument tube axial stack axial 2: Instrument tube thimble" region="-10004 10038" universe="10032" />
<cell id="10162" material="10015" name="Instrument tube axial stack axial 3: Water SPN" region="-10005 10004" universe="10032" />
<cell id="10163" material="10014" name="Instrument tube axial stack axial top: Borated Water" region="10005" universe="10032" />
<cell fill="10032" id="10164" name="(Instrument tube axial stack) wrapped by ((Empty Guide Tube in Center Position) wrapped by (Grids axial universe)) radial 0: Instrument tube axial stack" region="-10051" universe="10033" />
<cell fill="10029" id="10165" name="(Instrument tube axial stack) wrapped by ((Empty Guide Tube in Center Position) wrapped by (Grids axial universe)) radial outer: (Empty Guide Tube in Center Position) wrapped by (Grids axial universe)" region="10051" universe="10033" />
<cell id="10166" material="10001" name="BPRA rod lower/upper fitting radial 0: SS304" region="-10057" universe="10034" />
<cell id="10167" material="10014" name="BPRA rod lower/upper fitting radial outer: Borated Water" region="10057" universe="10034" />
<cell id="10168" material="10000" name="BPRA rod active poison radial 0: Air" region="-10052" universe="10035" />
<cell id="10169" material="10001" name="BPRA rod active poison radial 1: SS304" region="10052 -10053" universe="10035" />
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BIN
2x2-reflector/depleted/materials.xml (Stored with Git LFS)
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13
2x2-reflector/depleted/settings.xml
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@ -1,13 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
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11
2x2-reflector/depleted/tallies.xml
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244
2x2-reflector/fresh/geometry.xml
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@ -1,244 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
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</geometry>

164
2x2-reflector/fresh/materials.xml
View file

@ -1,164 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
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<density units="g/cc" value="10.30166" />
<nuclide ao="1.9992420000000022" name="O16" />
<nuclide ao="0.0007580000000000009" name="O17" />
<nuclide ao="0.0002523276152188021" name="U234" />
<nuclide ao="0.03140636057161555" name="U235" />
<nuclide ao="0.9683413118131656" name="U238" />
</material>
<material id="10013" name="Borosilicate Glass">
<temperature>300</temperature>
<density units="g/cc" value="2.26" />
<nuclide ao="0.013479369482225239" name="B10" />
<nuclide ao="0.054735873774104264" name="B11" />
<nuclide ao="0.6509787013744828" name="O16" />
<nuclide ao="0.00024681447050525047" name="O17" />
<nuclide ao="0.23640592474731761" name="Si28" />
<nuclide ao="0.01200401834354893" name="Si29" />
<nuclide ao="0.007913102535354443" name="Si30" />
<nuclide ao="0.024236195272461444" name="Al27" />
</material>
<material id="10014" name="Borated Water">
<temperature>300</temperature>
<density units="g/cc" value="0.7405820675158279" />
<nuclide ao="0.00032178659941803253" name="B10" />
<nuclide ao="0.0013017583017829388" name="B11" />
<nuclide ao="1.996441935899364" name="H1" />
<nuclide ao="0.0003109742982341739" name="H2" />
<nuclide ao="0.9979980704223166" name="O16" />
<nuclide ao="0.0003783846764824448" name="O17" />
<sab name="c_H_in_H2O" />
</material>
<material id="10015" name="Water SPN">
<density units="g/cc" value="0.9810025319057221" />
<nuclide ao="0.00032178659941803253" name="B10" />
<nuclide ao="0.0013017583017829388" name="B11" />
<nuclide ao="1.996441935899364" name="H1" />
<nuclide ao="0.0003109742982341739" name="H2" />
<nuclide ao="0.9979980704223166" name="O16" />
<nuclide ao="0.0003783846764824448" name="O17" />
<sab name="c_H_in_H2O" />
</material>
<material id="10016" name="SS SPN">
<temperature>300</temperature>
<density units="g/cc" value="3.6838480704877297" />
<nuclide name="Si28" wo="0.005512411036974801" />
<nuclide name="Si29" wo="0.0002899050175706448" />
<nuclide name="Si30" wo="0.00019768394545455493" />
<nuclide name="Cr50" wo="0.007930004298910168" />
<nuclide name="Cr52" wo="0.15902878860895933" />
<nuclide name="Cr53" wo="0.018379815074864116" />
<nuclide name="Cr54" wo="0.004661392017266364" />
<nuclide name="Mn55" wo="0.02" />
<nuclide name="Fe54" wo="0.03861561826636726" />
<nuclide name="Fe56" wo="0.6286064568062312" />
<nuclide name="Fe57" wo="0.014776921339264018" />
<nuclide name="Fe58" wo="0.002001003588137652" />
<nuclide name="Ni58" wo="0.06719770531879568" />
<nuclide name="Ni60" wo="0.02677596289274688" />
<nuclide name="Ni61" wo="0.0011833590846680462" />
<nuclide name="Ni62" wo="0.0038348222920813694" />
<nuclide name="Ni64" wo="0.0010081504117080411" />
</material>
</materials>

8
2x2-reflector/fresh/plots.xml
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<?xml version='1.0' encoding='utf-8'?>
<plots>
<plot basis="xy" color_by="material" filename="2x2-reflector" id="10020" type="slice">
<origin>0.0 0.0 195.0</origin>
<width>64.25184 64.25184</width>
<pixels>1000 1000</pixels>
</plot>
</plots>

15
2x2-reflector/fresh/settings.xml
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<?xml version='1.0' encoding='utf-8'?>
<settings>
<run_mode>eigenvalue</run_mode>
<particles>10000</particles>
<batches>10</batches>
<inactive>5</inactive>
<source strength="1.0">
<space type="fission">
<parameters>-32.12592 -32.12592 192.5 32.12592 32.12592 197.5</parameters>
</space>
</source>
<output>
<tallies>false</tallies>
</output>
</settings>

19
2x2-reflector/fresh/tallies.xml
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@ -1,19 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<tallies>
<filter id="1" type="distribcell">
<bins>10053</bins>
</filter>
<filter id="2" type="distribcell">
<bins>10061</bins>
</filter>
<tally id="10000" name="depletion tally">
<filters>1</filters>
<nuclides>O16 O17 U234 U235 U238</nuclides>
<scores>(n,p) (n,a) (n,gamma) fission (n,2n) (n,3n) (n,4n)</scores>
</tally>
<tally id="10001" name="depletion tally">
<filters>2</filters>
<nuclides>O16 O17 U234 U235 U238</nuclides>
<scores>(n,p) (n,a) (n,gamma) fission (n,2n) (n,3n) (n,4n)</scores>
</tally>
</tallies>

151
2x2-reflector/geometry.py
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import openmc
from beavrs.builder import BEAVRS
def find_assembly(assembly_name, wrap_geometry=True):
"""Find a fuel assembly with some string name in the BEAVRS OpenMC model.
This method extracts the fuel assembly and wraps it in an OpenMC Geometry.
The returned geometry has reflective boundary conditions along all
boundaries. The z-axis is bounded between z=200 and z=210 cm.
Parameters
----------
assembly_name : str
The name of the fuel assembly lattice
wrap_geometry : bool
If false, the fuel assembly Lattice is returned. If true, the fuel
assembly Lattice is wrapped in an OpenMC Geometry and returned (default).
Returns
-------
fuel_assembly
The OpenMC Lattice or Geometry for the assembly or None if not found
"""
# Get OpenMC Lattices for the fuel assembly
fuel_assembly = \
beavrs.openmc_geometry.get_lattices_by_name(assembly_name)[0]
# Wrap lattice in a Geometry if requested by the user
if wrap_geometry:
# Create a root Cell
root_cell = openmc.Cell(name='root cell')
root_cell.fill = fuel_assembly
# Make mixed reflective / vacuum boundaries
min_x = openmc.XPlane(x0=-10.70864, boundary_type='reflective')
max_x = openmc.XPlane(x0=+-10.70864, boundary_type='reflective')
min_y = openmc.YPlane(y0=-10.70864, boundary_type='reflective')
max_y = openmc.YPlane(y0=+-10.70864, boundary_type='reflective')
max_z = openmc.ZPlane(z0=197.5, boundary_type='reflective')
min_z = openmc.ZPlane(z0=192.5, boundary_type='reflective')
# Add boundaries to the root Cell
root_cell.region = \
+min_x & -max_x & +min_y & -max_y & +min_z & -max_z
# Create a root Universe
root_univ = openmc.Universe(universe_id=0, name='root universe')
root_univ.add_cell(root_cell)
# Create a Geometry
fuel_assembly = openmc.Geometry(root_univ)
return fuel_assembly
def build_reflector(assembly1_name, assembly2_name):
"""Build a 2x2 fuel assembly geometry with a water reflector on the
bottom and right.
This routine puts reflective boundary conditions along min x, max y and z
and vacuum boundary conditions along the max x and min y boundaries.
Parameters
----------
assembly1_name : str
The BEAVRS fuel assembly to place in the bottom right and top left
assembly2_name : str
The BEAVRS fuel assembly to place in the bottom left and top right
Returns
-------
openmc.Geometry
A 2x2 fuel assembly and reflector OpenMC Geometry
"""
fuel_assembly1 = find_assembly(assembly1_name, wrap_geometry=False)
fuel_assembly2 = find_assembly(assembly2_name, wrap_geometry=False)
# Find the water material
all_cells = beavrs.main_universe.get_all_cells()
for cell_uuid, cell in all_cells.items():
if cell.fill_type == 'material' and cell.fill.name == 'Borated Water':
water = cell.fill
# Create a Cell/Universe around the first fuel assembly
fuel_cell1 = openmc.Cell(name='assm1 cell')
fuel_cell1.fill = fuel_assembly1
fuel_univ1 = openmc.Universe(name='assm1 universe')
fuel_univ1.add_cell(fuel_cell1)
# Create a Cell/Universe around the second fuel assembly
fuel_cell2 = openmc.Cell(name='assm2 cell')
fuel_cell2.fill = fuel_assembly2
fuel_univ2 = openmc.Universe(name='assm2 universe')
fuel_univ2.add_cell(fuel_cell2)
# Create a Cell/Universe with water
water_cell = openmc.Cell(name='water cell', fill=water)
water_univ = openmc.Universe(name='water universe')
water_univ.add_cell(water_cell)
# Create a 3x3 lattice two fuel assemblies surrounded by a water reflector
reflector_lattice = openmc.RectLattice(name='reflector')
reflector_lattice.lower_left = [-32.12592, -32.12592, -500.]
reflector_lattice.pitch = [21.41728, 21.41728, 1000.]
reflector_lattice.universes = [[[fuel_univ1, fuel_univ2, water_univ],
[fuel_univ2, fuel_univ1, water_univ],
[water_univ, water_univ, water_univ]]]
# Create a Geometry around the reflected lattice
root_cell = openmc.Cell(name='root cell')
root_cell.fill = reflector_lattice
# Make mixed reflective / vacuum boundaries
min_x = openmc.XPlane(x0=-32.12592, boundary_type='reflective')
max_x = openmc.XPlane(x0=+32.12592, boundary_type='vacuum')
min_y = openmc.YPlane(y0=-32.12592, boundary_type='vacuum')
max_y = openmc.YPlane(y0=+32.12592, boundary_type='reflective')
min_z = openmc.ZPlane(z0=192.5, boundary_type='reflective')
max_z = openmc.ZPlane(z0=197.5, boundary_type='reflective')
# Add boundaries to the root Cell
root_cell.region = +min_x & -max_x & +min_y & -max_y & +min_z & -max_z
# Create a root Universe for this fuel assembly
root_univ = openmc.Universe(universe_id=0, name='root universe')
root_univ.add_cell(root_cell)
# Create an OpenMC Geometry for this fuel assembly
reflector = openmc.Geometry()
reflector.root_universe = root_univ
return reflector
#### Create OpenMC "materials.xml" and "geometry.xml" files
# Instantiate a BEAVRS object
beavrs = BEAVRS()
# Write all BEAVRS materials to materials.xml file
beavrs.write_openmc_materials()
# Extract fuel assemblies of interest from BEAVRS model
openmc_geometry = build_reflector('Fuel 1.6% enr instr no BAs',
'Fuel 3.1% enr instr 20')

54
README.md
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@ -1,43 +1,37 @@
ECP Benchmarks
==============
ExaSMR Benchmarks
=================
This repository contains benchmarks for performance profiling of the OpenMC
and Shift Monte Carlo codes for ECP. Each directory contains a ``build-xml.py``
This repository contains benchmarks for performance profiling of the OpenMC and
Shift Monte Carlo codes for the ECP ExaSMR project. Each directory contains a
Python script which uses the OpenMC Python API to create XML input files for
OpenMC. The Python scripts and the XML files are included in this repository.
Each benchmark is derived from a subset (*e.g.*, fuel pins, assemblies) of the
full-core BEAVRS model with fresh UO2 fuel at HZP conditions. Each benchmark is
modeled in 2D with reflective boundary conditions along the axial dimension.
OpenMC.
Model Descriptions
------------------
<dl>
<dt>fuel-pin</dt>
<dd>A single fuel pin with reflective boundary conditions (i.e., an infinitely repeating array of fuel pins). The default configuration is the BEAVRS fuel pin with fresh 1.6% enriched UO2 fuel, but the script can be toggled to 2.4% or 3.1% enriched fuel.</dd>
<dt>assembly</dt>
<dd>A single fuel assembly with reflective boundary conditions (i.e., an infinitely repeating lattice of fuel assemblies). The default configuration is the BEAVRS assembly with fresh 1.6% enriched UO2 fuel with 24 water-filled control rod guide tubes and a central air-filled instrument tube. However, the script can be toggled to use any of the 20+ fuel assemblies in the BEAVRS model.</dd>
<dt>2x2-periodic</dt>
<dd>A 2x2 fuel assembly colorset with periodic boundary conditions (i.e., an infinitely repeating lattice of the 2x2 assembly colorset). The default configuration includes the BEAVRS assembly with fresh 1.6% enriched UO2 fuel with 24 water-filled control rod guide tubes and a central air-filled instrument tube, along with a 3.1% enriched fuel assembly with 20 burnable poisons, four control rod guide tubes and a central instrument tube. However, the script may be toggled to use any pair of the 20+ fuel assemblies in the BEAVRS model.</dd>
<dt>2x2-reflector</dt>
<dd>A 2x2 fuel assembly colorset surrounded by a water reflector. Reflective boundary conditions are used on the top and left boundaries (adjacent to the assemblies) and vacuum boundary conditions are used on the bottom and right boundaries (adjacent to the reflector). The default configuration includes the BEAVRS assembly with fresh 1.6% enriched UO2 fuel with 24 water-filled control rod guide tubes and a central air-filled instrument tube, along with a 3.1% enriched fuel assembly with 20 burnable poisons, four control rod guide tubes and a central instrument tube. However, the script may be toggled to use any pair of the 20+ fuel assemblies in the BEAVRS model.</dd>
<dt>smr</dt>
<dd>A 3D Small Modular Reactor (SMR) model that roughly mimics the design of the NuScale reactor. The reactor core has 37 fuel assemblies that alternate between 3.1% enriched and 2.4% enriched UO2 fuel. The center fuel assembly has 1.6% enriched UO2 fuel. Where possible, we have attempted to use the same parameters for the fuel assemblies and fuel rods as those specified in the NuScale <a href="https://www.nrc.gov/reactors/new-reactors/design-cert/nuscale.html">design submittal</a> to the Nuclear Regulatory Commission (NRC), for example: 264 fuel rods per assembly, 24 guide tubes per assembly, 1 instrument tube per assembly, five spacer grids per assembly, fuel rod pitch of 0.496 in, and an active fuel length of 200 cm. Many of the details of the actual NuScale fuel assembly design are redacted from the design submittal because they are export controlled/proprietary information. The purpose of our model is not to be an exact replica of the NuScale model; rather, it is intended to capture most of the physical complexities that are involved in modeling a full reactor core and to provide a suitable model for carrying out full core performance tests on the testbed architectures.</dd>
<dd>A 3D Small Modular Reactor (SMR) model that roughly mimics the design of
the NuScale reactor. The reactor core has 37 fuel assemblies that alternate
between 3.1% enriched and 2.4% enriched UO2 fuel. The center fuel assembly
has 1.6% enriched UO2 fuel. Where possible, we have attempted to use the
same parameters for the fuel assemblies and fuel rods as those specified in
the NuScale <a
href="https://www.nrc.gov/reactors/new-reactors/design-cert/nuscale.html">design
submittal</a> to the Nuclear Regulatory Commission (NRC), for example: 264
fuel rods per assembly, 24 guide tubes per assembly, 1 instrument tube per
assembly, five spacer grids per assembly, fuel rod pitch of 0.496 in, and an
active fuel length of 200 cm. Many of the details of the actual NuScale fuel
assembly design are redacted from the design submittal because they are
export controlled/proprietary information. The purpose of our model is not
to be an exact replica of the NuScale model; rather, it is intended to
capture most of the physical complexities that are involved in modeling a
full reactor core and to provide a suitable model for carrying out full core
performance tests on the testbed architectures.</dd>
</dl>
Dependencies
------------
These scripts depend on the following Python packages:
* mit-crpg/openmc (develop branch)
* mit-crpg/PWR_benchmarks
Of particular note, The Python ``beavrs`` package must be installed from the
mit-crpg/PWR_benchmarks repository. This package may be installed with
``distutils`` as follows:
```bash
cd PWR_benchmarks/BEAVRS/openmc/inputs
python setup.py install
```
To generate models using the scripts in this package, you must use a recent
version of OpenMC's Python API.

59
assembly/build-depleted.py
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#!/usr/bin/env python3
import numpy as np
import opendeplete
from geometry import openmc_geometry
# FIXME: Automatically extract info needed to calculate burnable cell volumes
# Fuel rod geometric parameters
radius = 0.39218
height = 5.
# Count the number of instances for each cell and material
openmc_geometry.determine_paths(instances_only=True)
# Extract all cells filled by a fuel material
fuel_cells = openmc_geometry.get_cells_by_name(
name='enr radial 0: Fuel', case_sensitive=True)
# Assign distribmats for each material
for cell in fuel_cells:
cell.fill.volume = np.pi * radius**2 * height
cell.fill.depletable = True
cell.fill.temperature = 300.0
cell.fill = [cell.fill.clone() for i in range(cell.num_instances)]
# Set temperature for all cells
cells = openmc_geometry.get_all_cells()
for cell_id in cells:
cells[cell_id].temperature = 300.0
# Create dt vector for 1 month with 5 day timesteps
dt1 = 5*24*60*60 # 5 days
dt2 = 1.*30*24*60*60 # 1 months
N = np.floor(dt2/dt1)
dt = np.repeat([dt1], N)
# Create settings variable
settings = opendeplete.OpenMCSettings()
settings.openmc_call = "openmc"
settings.particles = 30000
settings.batches = 20
settings.inactive = 10
settings.lower_left = [-10.70864, -10.70864, +192.5]
settings.upper_right = [+10.70864, +10.70864, +197.5]
settings.entropy_dimension = [17, 17, 1]
# MeV/second cm from CASMO
settings.power = 2.337e15 * (17.**2 / 1.5**2) * height
settings.dt_vec = dt
settings.output_dir = 'depleted'
op = opendeplete.OpenMCOperator(openmc_geometry, settings)
# Perform simulation using the MCNPX/MCNP6 algorithm
opendeplete.cecm(op)

140
assembly/build-fresh.py
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#!/usr/bin/env python3
"""Creates a 2D fuel assembly with reflective BCs."""
import os
import shutil
import copy
import numpy as np
import openmc
from geometry import beavrs, openmc_geometry
#### Query the user for options
# Query the user on whether to use multipole cross sections
multipole = input('Use multipole cross sections? (y/n): ').lower()
multipole = (multipole == 'y')
# Query the user on whether to use distribmats or distribcells
# If using distribmats, the geometry must be "differentiated" with unique
# material instances for each instance of a fuel cell
distrib = input('Use distribmat or distribcells? [mat/cell]: ').lower()
if distrib not in ['cell', 'mat']:
raise InputError('Distrib type "{}" is unsupported'.format(distrib))
#### "Differentiate" the geometry if using distribmats
if distrib == 'mat':
# Count the number of instances for each cell and material
openmc_geometry.determine_paths()
# Extract all cells filled by a fuel material
fuel_cells = openmc_geometry.get_cells_by_name(
name='enr radial 0: Fuel', case_sensitive=True)
# Assign distribmats for each material
for cell in fuel_cells:
new_materials = []
for i in range(cell.num_instances):
new_materials.append(cell.fill.clone())
# Fill cell with list of "differentiated" materials
cell.fill = new_materials
#### Create OpenMC "materials.xml" file
all_materials = openmc_geometry.get_all_materials()
materials = openmc.Materials(all_materials.values())
materials.export_to_xml()
#### Create OpenMC "geometry.xml" file
openmc_geometry.export_to_xml()
#### Create OpenMC "settings.xml" file
# Construct uniform initial source distribution over fissionable zones
lower_left = [-10.70864, -10.70864, +192.5]
upper_right = [+10.70864, +10.70864, +197.5]
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
source.space.only_fissionable = True
settings_file = openmc.Settings()
settings_file.batches = 10
settings_file.inactive = 5
settings_file.particles = 10000
settings_file.output = {'tallies': False}
settings_file.source = source
settings_file.sourcepoint_write = False
if multipole:
settings_file.temperature = {'multipole': True, 'tolerance': 1000}
settings_file.export_to_xml()
#### Create OpenMC "plots.xml" file
# Initialize the BEAVRS color mapping scheme
beavrs.write_openmc_plots()
# Create a plot colored by materials
plot = openmc.Plot()
plot.width = [10.70864*2, 10.70864*2]
plot.origin = [0., 0., 195.]
plot.color_by = 'material'
plot.filename = 'assembly'
plot.pixels = [1000, 1000]
plot_file = openmc.Plots([plot])
plot_file.export_to_xml()
#### Create OpenMC "tallies.xml" file
tallies = openmc.Tallies()
# Extract all fuel materials
materials = openmc_geometry.get_materials_by_name(name='Fuel', matching=False)
# If using distribcells, create distribcell tally needed for depletion
if distrib == 'cell':
fuel_cells = openmc_geometry.get_cells_by_name(
name='enr radial 0: Fuel', case_sensitive=True, matching=False)
for cell in fuel_cells:
tally = openmc.Tally(name='depletion tally')
tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
'fission', '(n,2n)', '(n,3n)', '(n,4n)']
tally.nuclides = cell.fill.get_nuclides()
tally.filters.append(openmc.DistribcellFilter([cell.id]))
tallies.append(tally)
# If using distribmats, create material tally needed for depletion
elif distrib == 'mat':
tally = openmc.Tally(name='depletion tally')
tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
'fission', '(n,2n)', '(n,3n)', '(n,4n)']
tally.nuclides = materials[0].get_nuclides()
material_ids = [material.id for material in materials]
tally.filters.append(openmc.MaterialFilter(material_ids))
tallies.append(tally)
tallies.export_to_xml()
#### Move all XML files to 'fresh' directory
if not os.path.exists('fresh'):
os.makedirs('fresh')
shutil.move('materials.xml', 'fresh/materials.xml')
shutil.move('geometry.xml', 'fresh/geometry.xml')
shutil.move('settings.xml', 'fresh/settings.xml')
shutil.move('tallies.xml', 'fresh/tallies.xml')
shutil.move('plots.xml', 'fresh/plots.xml')

159
assembly/depleted/geometry.xml
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<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="10001" material="10014" name="Intermediate grid pincell radial 0: Borated Water" region="10010 -10011 10012 -10013" universe="10001" />
<cell id="10002" material="10006" name="Intermediate grid pincell radial outer: Zircaloy 4" region="~(10010 -10011 10012 -10013)" universe="10001" />
<cell id="10003" material="10014" name="Top/Bottom grid pincell radial 0: Borated Water" region="10006 -10007 10008 -10009" universe="10002" />
<cell id="10004" material="10005" name="Top/Bottom grid pincell radial outer: Inconel 718" region="~(10006 -10007 10008 -10009)" universe="10002" />
<cell id="10009" material="10014" name="Grids axial universe axial 0: Borated Water" region="-10000" universe="10005" />
<cell id="10010" material="10015" name="Grids axial universe axial 1: Water SPN" region="-10003 10000" universe="10005" />
<cell id="10011" material="10014" name="Grids axial universe axial 2: Borated Water" region="-10018 10003" universe="10005" />
<cell fill="10002" id="10012" name="Grids axial universe axial 3: Top/Bottom grid pincell" region="-10019 10018" universe="10005" />
<cell id="10013" material="10014" name="Grids axial universe axial 4: Borated Water" region="-10020 10019" universe="10005" />
<cell fill="10001" id="10014" name="Grids axial universe axial 5: Intermediate grid pincell" region="-10021 10020" universe="10005" />
<cell id="10015" material="10014" name="Grids axial universe axial 6: Borated Water" region="-10022 10021" universe="10005" />
<cell fill="10001" id="10016" name="Grids axial universe axial 7: Intermediate grid pincell" region="-10023 10022" universe="10005" />
<cell id="10017" material="10014" name="Grids axial universe axial 8: Borated Water" region="-10024 10023" universe="10005" />
<cell fill="10001" id="10018" name="Grids axial universe axial 9: Intermediate grid pincell" region="-10025 10024" universe="10005" />
<cell id="10019" material="10014" name="Grids axial universe axial 10: Borated Water" region="-10026 10025" universe="10005" />
<cell fill="10001" id="10020" name="Grids axial universe axial 11: Intermediate grid pincell" region="-10027 10026" universe="10005" />
<cell id="10021" material="10014" name="Grids axial universe axial 12: Borated Water" region="-10028 10027" universe="10005" />
<cell fill="10001" id="10022" name="Grids axial universe axial 13: Intermediate grid pincell" region="-10029 10028" universe="10005" />
<cell id="10023" material="10014" name="Grids axial universe axial 14: Borated Water" region="-10030 10029" universe="10005" />
<cell fill="10001" id="10024" name="Grids axial universe axial 15: Intermediate grid pincell" region="-10031 10030" universe="10005" />
<cell id="10025" material="10014" name="Grids axial universe axial 16: Borated Water" region="-10032 10031" universe="10005" />
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BIN
assembly/depleted/materials.xml (Stored with Git LFS)
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Binary file not shown.

13
assembly/depleted/settings.xml
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@ -1,13 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<settings>
<run_mode>eigenvalue</run_mode>
<particles>30000</particles>
<batches>20</batches>
<inactive>10</inactive>
<source strength="1.0">
<space type="box">
<parameters>-10.70864 -10.70864 192.5 10.70864 10.70864 197.5</parameters>
</space>
</source>
<seed>7238470365726585870</seed>
</settings>

11
assembly/depleted/tallies.xml
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@ -1,11 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<tallies>
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</tally>
</tallies>

159
assembly/fresh/geometry.xml
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@ -1,159 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
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<cell id="10025" material="10014" name="Grids axial universe axial 16: Borated Water" region="-10032 10031" universe="10005" />
<cell fill="10002" id="10026" name="Grids axial universe axial 17: Top/Bottom grid pincell" region="-10033 10032" universe="10005" />
<cell id="10027" material="10014" name="Grids axial universe axial 18: Borated Water" region="-10004 10033" universe="10005" />
<cell id="10028" material="10015" name="Grids axial universe axial 19: Water SPN" region="-10005 10004" universe="10005" />
<cell id="10029" material="10014" name="Grids axial universe axial top: Borated Water" region="10005" universe="10005" />
<cell id="10051" material="10006" name="Fuel rod lower/upper fitting radial 0: Zircaloy 4" region="-10036" universe="10007" />
<cell id="10052" material="10014" name="Fuel rod lower/upper fitting radial outer: Borated Water" region="10036" universe="10007" />
<cell id="10053" material="10008" name="Fuel rod active region - 1.6% enr radial 0: Fuel 1.6%" region="-10034" universe="10008" />
<cell id="10054" material="10004" name="Fuel rod active region - 1.6% enr radial 1: Helium" region="10034 -10035" universe="10008" />
<cell id="10055" material="10006" name="Fuel rod active region - 1.6% enr radial 2: Zircaloy 4" region="10035 -10036" universe="10008" />
<cell id="10056" material="10014" name="Fuel rod active region - 1.6% enr radial outer: Borated Water" region="10036" universe="10008" />
<cell id="10073" material="10005" name="Fuel rod plenum radial 0: Inconel 718" region="-10037" universe="10013" />
<cell id="10074" material="10004" name="Fuel rod plenum radial 1: Helium" region="10037 -10035" universe="10013" />
<cell id="10075" material="10006" name="Fuel rod plenum radial 2: Zircaloy 4" region="10035 -10036" universe="10013" />
<cell id="10076" material="10014" name="Fuel rod plenum radial outer: Borated Water" region="10036" universe="10013" />
<cell id="10077" material="10014" name="Fuel rod - 1.6% enr axial 0: Borated Water" region="-10000" universe="10014" />
<cell id="10078" material="10016" name="Fuel rod - 1.6% enr axial 1: SS SPN" region="-10038 10000" universe="10014" />
<cell fill="10007" id="10079" name="Fuel rod - 1.6% enr axial 2: Fuel rod lower/upper fitting" region="-10039 10038" universe="10014" />
<cell fill="10008" id="10080" name="Fuel rod - 1.6% enr axial 3: Fuel rod active region - 1.6% enr" region="-10040 10039" universe="10014" />
<cell fill="10013" id="10081" name="Fuel rod - 1.6% enr axial 4: Fuel rod plenum" region="-10041 10040" universe="10014" />
<cell fill="10007" id="10082" name="Fuel rod - 1.6% enr axial 5: Fuel rod lower/upper fitting" region="-10042 10041" universe="10014" />
<cell id="10083" material="10014" name="Fuel rod - 1.6% enr axial 6: Borated Water" region="-10004 10042" universe="10014" />
<cell id="10084" material="10016" name="Fuel rod - 1.6% enr axial 7: SS SPN" region="-10005 10004" universe="10014" />
<cell id="10085" material="10014" name="Fuel rod - 1.6% enr axial top: Borated Water" region="10005" universe="10014" />
<cell fill="10014" id="10086" name="(Fuel rod - 1.6% enr) wrapped by (Grids axial universe) radial 0: Fuel rod - 1.6% enr" region="-10036" universe="10015" />
<cell fill="10005" id="10087" name="(Fuel rod - 1.6% enr) wrapped by (Grids axial universe) radial outer: Grids axial universe" region="10036" universe="10015" />
<cell id="10132" material="10014" name="Empty GT below the dashpot radial 0: Borated Water" region="-10045" universe="10024" />
<cell id="10133" material="10006" name="Empty GT below the dashpot radial 1: Zircaloy 4" region="10045 -10046" universe="10024" />
<cell id="10134" material="10014" name="Empty GT below the dashpot radial outer: Borated Water" region="10046" universe="10024" />
<cell id="10135" material="10014" name="Empty GT above the dashpot radial 0: Borated Water" region="-10043" universe="10025" />
<cell id="10136" material="10006" name="Empty GT above the dashpot radial 1: Zircaloy 4" region="10043 -10044" universe="10025" />
<cell id="10137" material="10014" name="Empty GT above the dashpot radial outer: Borated Water" region="10044" universe="10025" />
<cell id="10138" material="10014" name="Empty Guide Tube axial 0: Borated Water" region="-10000" universe="10026" />
<cell id="10139" material="10015" name="Empty Guide Tube axial 1: Water SPN" region="-10047 10000" universe="10026" />
<cell fill="10024" id="10140" name="Empty Guide Tube axial 2: Empty GT below the dashpot" region="-10048 10047" universe="10026" />
<cell fill="10025" id="10141" name="Empty Guide Tube axial 3: Empty GT above the dashpot" region="-10049 10048" universe="10026" />
<cell id="10142" material="10015" name="Empty Guide Tube axial 4: Water SPN" region="-10005 10049" universe="10026" />
<cell id="10143" material="10014" name="Empty Guide Tube axial top: Borated Water" region="10005" universe="10026" />
<cell fill="10026" id="10144" name="(Empty Guide Tube) wrapped by (Grids axial universe) radial 0: Empty Guide Tube" region="-10044" universe="10027" />
<cell fill="10005" id="10145" name="(Empty Guide Tube) wrapped by (Grids axial universe) radial outer: Grids axial universe" region="10044" universe="10027" />
<cell id="10146" material="10014" name="Empty Guide Tube in Center Position axial 0: Borated Water" region="-10000" universe="10028" />
<cell id="10147" material="10015" name="Empty Guide Tube in Center Position axial 1: Water SPN" region="-10047 10000" universe="10028" />
<cell fill="10025" id="10148" name="Empty Guide Tube in Center Position axial 2: Empty GT above the dashpot" region="-10049 10047" universe="10028" />
<cell id="10149" material="10015" name="Empty Guide Tube in Center Position axial 3: Water SPN" region="-10005 10049" universe="10028" />
<cell id="10150" material="10014" name="Empty Guide Tube in Center Position axial top: Borated Water" region="10005" universe="10028" />
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<cell fill="10005" id="10152" name="(Empty Guide Tube in Center Position) wrapped by (Grids axial universe) radial outer: Grids axial universe" region="10044" universe="10029" />
<cell id="10153" material="10000" name="Instrument tube thimble radial 0: Air" region="-10050" universe="10030" />
<cell id="10154" material="10006" name="Instrument tube thimble radial 1: Zircaloy 4" region="10050 -10051" universe="10030" />
<cell id="10155" material="10014" name="Instrument tube thimble radial outer: Borated Water" region="10051" universe="10030" />
<cell id="10156" material="10000" name="Instrument tube thimble support plane radial 0: Air" region="-10050" universe="10031" />
<cell id="10157" material="10006" name="Instrument tube thimble support plane radial 1: Zircaloy 4" region="10050 -10051" universe="10031" />
<cell id="10158" material="10015" name="Instrument tube thimble support plane radial outer: Water SPN" region="10051" universe="10031" />
<cell fill="10030" id="10159" name="Instrument tube axial stack axial 0: Instrument tube thimble" region="-10000" universe="10032" />
<cell fill="10031" id="10160" name="Instrument tube axial stack axial 1: Instrument tube thimble support plane" region="-10038 10000" universe="10032" />
<cell fill="10030" id="10161" name="Instrument tube axial stack axial 2: Instrument tube thimble" region="-10004 10038" universe="10032" />
<cell id="10162" material="10015" name="Instrument tube axial stack axial 3: Water SPN" region="-10005 10004" universe="10032" />
<cell id="10163" material="10014" name="Instrument tube axial stack axial top: Borated Water" region="10005" universe="10032" />
<cell fill="10032" id="10164" name="(Instrument tube axial stack) wrapped by ((Empty Guide Tube in Center Position) wrapped by (Grids axial universe)) radial 0: Instrument tube axial stack" region="-10051" universe="10033" />
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<cell fill="10065" id="11273" name="root cell" region="10149 -10150 10151 -10152 10154 -10153" universe="0" />
<lattice id="10065" name="Fuel 1.6% enr instr no BAs">
<pitch>1.25984 1.25984</pitch>
<dimension>17 17</dimension>
<lower_left>-10.70864 -10.70864</lower_left>
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10015 10015 10015 10015 10015 10015 10015 10015 10015 10015 10015 10015 10015 10015 10015 10015 10015
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10015 10015 10015 10015 10015 10015 10015 10015 10015 10015 10015 10015 10015 10015 10015 10015 10015 </universes>
</lattice>
<surface coeffs="20.0" id="10000" name="Support plate bottom" type="z-plane" />
<surface coeffs="35.0" id="10003" name="Lower nozzle top" type="z-plane" />
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<surface coeffs="431.876" id="10005" name="Upper nozzle top" type="z-plane" />
<surface coeffs="-0.61015" id="10006" name="minimum x" type="x-plane" />
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<surface boundary="reflective" coeffs="192.5" id="10154" type="z-plane" />
</geometry>

122
assembly/fresh/materials.xml
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@ -1,122 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<materials>
<material id="10000" name="Air">
<temperature>300</temperature>
<density units="g/cc" value="0.00616" />
<nuclide ao="0.2094205995" name="O16" />
<nuclide ao="7.94005e-05" name="O17" />
<nuclide ao="0.7780395633" name="N14" />
<nuclide ao="0.0028604367000000003" name="N15" />
<nuclide ao="3.1124879999999996e-05" name="Ar36" />
<nuclide ao="5.86857e-06" name="Ar38" />
<nuclide ao="0.00929300655" name="Ar40" />
<nuclide ao="0.00027" name="C0" />
</material>
<material id="10004" name="Helium">
<temperature>300</temperature>
<density units="g/cc" value="0.0015981" />
<nuclide name="He3" wo="1.5070346049256974e-06" />
<nuclide name="He4" wo="0.999998492965395" />
</material>
<material id="10005" name="Inconel 718">
<temperature>300</temperature>
<density units="g/cc" value="8.2" />
<nuclide name="Si28" wo="0.003215573104901967" />
<nuclide name="Si29" wo="0.00016911126024954278" />
<nuclide name="Si30" wo="0.00011531563484849038" />
<nuclide name="Cr50" wo="0.007913309553017726" />
<nuclide name="Cr52" wo="0.15869399115925625" />
<nuclide name="Cr53" wo="0.018341120727338085" />
<nuclide name="Cr54" wo="0.004651578560387908" />
<nuclide name="Mn55" wo="0.0087" />
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<nuclide name="Fe56" wo="0.26311407687664323" />
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</material>
<material id="10006" name="Zircaloy 4">
<temperature>300</temperature>
<density units="g/cc" value="6.55" />
<nuclide name="O16" wo="0.0012494965182849112" />
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<material id="10008" name="Fuel 1.6%">
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<material id="10014" name="Borated Water">
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<material id="10015" name="Water SPN">
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<material id="10016" name="SS SPN">
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</material>
</materials>

8
assembly/fresh/plots.xml
View file

@ -1,8 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<plots>
<plot basis="xy" color_by="material" filename="assembly" id="10020" type="slice">
<origin>0.0 0.0 195.0</origin>
<width>21.41728 21.41728</width>
<pixels>1000 1000</pixels>
</plot>
</plots>

15
assembly/fresh/settings.xml
View file

@ -1,15 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<settings>
<run_mode>eigenvalue</run_mode>
<particles>10000</particles>
<batches>10</batches>
<inactive>5</inactive>
<source strength="1.0">
<space type="fission">
<parameters>-10.70864 -10.70864 192.5 10.70864 10.70864 197.5</parameters>
</space>
</source>
<output>
<tallies>false</tallies>
</output>
</settings>

11
assembly/fresh/tallies.xml
View file

@ -1,11 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<tallies>
<filter id="1" type="distribcell">
<bins>10053</bins>
</filter>
<tally id="10000" name="depletion tally">
<filters>1</filters>
<nuclides>O16 O17 U234 U235 U238</nuclides>
<scores>(n,p) (n,a) (n,gamma) fission (n,2n) (n,3n) (n,4n)</scores>
</tally>
</tallies>

65
assembly/geometry.py
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@ -1,65 +0,0 @@
import openmc
from beavrs.builder import BEAVRS
def find_assembly(assembly_name, wrap_geometry=True):
"""Find a fuel assembly with some string name in the BEAVRS OpenMC model.
This method extracts the fuel assembly and wraps it in an OpenMC Geometry.
The returned geometry has reflective boundary conditions along all
boundaries. The z-axis is bounded between z=200 and z=210 cm.
Parameters
----------
assembly_name : str
The name of the fuel assembly lattice
wrap_geometry : bool
If false, the fuel assembly Lattice is returned. If true, the fuel
assembly Lattice is wrapped in an OpenMC Geometry and returned (default).
Returns
-------
fuel_assembly
The OpenMC Lattice or Geometry for the assembly or None if not found
"""
# Get OpenMC Lattices for the fuel assembly
fuel_assembly = \
beavrs.openmc_geometry.get_lattices_by_name(assembly_name)[0]
# Wrap lattice in a Geometry if requested by the user
if wrap_geometry:
# Create a root Cell
root_cell = openmc.Cell(name='root cell')
root_cell.fill = fuel_assembly
# Make mixed reflective / vacuum boundaries
min_x = openmc.XPlane(x0=-10.70864, boundary_type='reflective')
max_x = openmc.XPlane(x0=+10.70864, boundary_type='reflective')
min_y = openmc.YPlane(y0=-10.70864, boundary_type='reflective')
max_y = openmc.YPlane(y0=+10.70864, boundary_type='reflective')
max_z = openmc.ZPlane(z0=197.5, boundary_type='reflective')
min_z = openmc.ZPlane(z0=192.5, boundary_type='reflective')
# Add boundaries to the root Cell
root_cell.region = \
+min_x & -max_x & +min_y & -max_y & +min_z & -max_z
# Create a root Universe
root_univ = openmc.Universe(universe_id=0, name='root universe')
root_univ.add_cell(root_cell)
# Create a Geometry
fuel_assembly = openmc.Geometry(root_univ)
return fuel_assembly
# Instantiate a BEAVRS object
beavrs = BEAVRS()
# Extract fuel assembly of interest from BEAVRS model
assm_name = 'Fuel 1.6% enr instr no BAs'
openmc_geometry = find_assembly(assm_name)

59
pin-cell/build-depleted.py
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@ -1,59 +0,0 @@
#!/usr/bin/env python3
import numpy as np
import opendeplete
from geometry import openmc_geometry
# FIXME: Automatically extract info needed to calculate burnable cell volumes
# Fuel rod geometric parameters
radius = 0.39218
height = 5.
# Count the number of instances for each cell and material
openmc_geometry.determine_paths(instances_only=True)
# Extract all cells filled by a fuel material
fuel_cells = openmc_geometry.get_cells_by_name(
name='enr radial 0: Fuel', case_sensitive=True)
# Assign distribmats for each material
for cell in fuel_cells:
cell.fill.volume = np.pi * radius**2 * height
cell.fill.depletable = True
cell.fill.temperature = 300.0
cell.fill = [cell.fill.clone() for i in range(cell.num_instances)]
# Set temperature for all cells
cells = openmc_geometry.get_all_cells()
for cell_id in cells:
cells[cell_id].temperature = 300.0
# Create dt vector for 1 month with 5 day timesteps
dt1 = 5*24*60*60 # 5 days
dt2 = 1.*30*24*60*60 # 1 months
N = np.floor(dt2/dt1)
dt = np.repeat([dt1], N)
# Create settings variable
settings = opendeplete.OpenMCSettings()
settings.openmc_call = "openmc"
settings.particles = 10000
settings.batches = 20
settings.inactive = 10
settings.lower_left = [-0.62992, -0.62992, -10.0]
settings.upper_right = [+0.62992, +0.62992, +10.0]
settings.entropy_dimension = [1, 1, 1]
# MeV/second cm from CASMO
settings.power = 2.337e15 * (1.**2 / 1.5**2) * height
settings.dt_vec = dt
settings.output_dir = 'depleted'
op = opendeplete.OpenMCOperator(openmc_geometry, settings)
# Perform simulation using the MCNPX/MCNP6 algorithm
opendeplete.cecm(op)

98
pin-cell/build-fresh.py
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@ -1,98 +0,0 @@
#!/usr/bin/env python3
"""Creates a 2D fuel pin with reflective BCs."""
import os
import shutil
import numpy as np
import openmc
from geometry import beavrs, openmc_geometry
#### Query the user for options
# Query the user on whether to use multipole cross sections
multipole = input('Use multipole cross sections? (y/n): ').lower()
multipole = (multipole == 'y')
#### Create OpenMC "materials.xml" file
beavrs.write_openmc_materials()
#### Create OpenMC "geometry.xml" file
openmc_geometry.export_to_xml()
#### Create OpenMC "settings.xml" file
# Construct uniform initial source distribution over fissionable zones
lower_left = [-0.62992, -0.62992, -10.0]
upper_right = [+0.62992, +0.62992, +10.0]
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
source.space.only_fissionable = True
settings_file = openmc.Settings()
settings_file.batches = 10
settings_file.inactive = 5
settings_file.particles = 10000
settings_file.output = {'tallies': False}
settings_file.source = source
settings_file.sourcepoint_write = False
if multipole:
settings_file.temperature = {'multipole': True, 'tolerance': 1000}
settings_file.export_to_xml()
#### Create OpenMC "plots.xml" file
# Initialize the BEAVRS color mapping scheme
beavrs.write_openmc_plots()
# Create a plot colored by materials
plot = openmc.Plot()
plot.width = [1.25984, 1.25984]
plot.origin = [0., 0., np.inf]
plot.color_by = 'material'
plot.filename = 'fuel-pin'
plot.colors = beavrs.plots.colors_mat
plot.pixels = [1000, 1000]
plot_file = openmc.Plots([plot])
plot_file.export_to_xml()
#### Create a tally akin to that used by OpenDeplete for depletion
# Extract all fuel materials
materials = openmc_geometry.get_materials_by_name(name='Fuel', matching=False)
# If using distribmats, create material tally needed for depletion
tally = openmc.Tally(name='depletion tally')
tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
'fission', '(n,2n)', '(n,3n)', '(n,4n)']
tally.nuclides = materials[0].get_nuclides()
material_ids = [material.id for material in materials]
tally.filters.append(openmc.MaterialFilter(material_ids))
#### Create OpenMC "tallies.xml" file
tallies = openmc.Tallies()
tallies.append(tally)
tallies.export_to_xml()
#### Move all XML files to 'fresh' directory
if not os.path.exists('fresh'):
os.makedirs('fresh')
shutil.move('materials.xml', 'fresh/materials.xml')
shutil.move('geometry.xml', 'fresh/geometry.xml')
shutil.move('settings.xml', 'fresh/settings.xml')
shutil.move('tallies.xml', 'fresh/tallies.xml')
shutil.move('plots.xml', 'fresh/plots.xml')

15
pin-cell/depleted/geometry.xml
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@ -1,15 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="10053" material="10015" name="Fuel rod active region - 1.6% enr radial 0: Fuel 1.6%" region="-10034" universe="10008" />
<cell id="10054" material="10004" name="Fuel rod active region - 1.6% enr radial 1: Helium" region="10034 -10035" universe="10008" />
<cell id="10055" material="10006" name="Fuel rod active region - 1.6% enr radial 2: Zircaloy 4" region="10035 -10036" universe="10008" />
<cell id="10056" material="10014" name="Fuel rod active region - 1.6% enr radial outer: Borated Water" region="10036" universe="10008" />
<cell fill="10008" id="11273" name="root cell" region="10149 -10150 10151 -10152" universe="0" />
<surface coeffs="0.0 0.0 0.39218" id="10034" name="Fuel pellet OR" type="z-cylinder" />
<surface coeffs="0.0 0.0 0.40005" id="10035" name="Fuel clad IR" type="z-cylinder" />
<surface coeffs="0.0 0.0 0.4572" id="10036" name="Fuel clad OR" type="z-cylinder" />
<surface boundary="reflective" coeffs="-0.62992" id="10149" type="x-plane" />
<surface boundary="reflective" coeffs="0.62992" id="10150" type="x-plane" />
<surface boundary="reflective" coeffs="-0.62992" id="10151" type="y-plane" />
<surface boundary="reflective" coeffs="0.62992" id="10152" type="y-plane" />
</geometry>

343
pin-cell/depleted/materials.xml
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@ -1,343 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<materials>
<material id="10004">
<temperature>300</temperature>
<density units="sum" />
<nuclide ao="4.80886434994e-10" name="He3" />
<nuclide ao="0.000240442736611" name="He4" />
</material>
<material id="10006">
<temperature>300</temperature>
<density units="sum" />
<nuclide ao="0.00030813805257" name="O16" />
<nuclide ao="1.16828599964e-07" name="O17" />
<nuclide ao="3.29618026316e-06" name="Cr50" />
<nuclide ao="6.35635553671e-05" name="Cr52" />
<nuclide ao="7.20759693447e-06" name="Cr53" />
<nuclide ao="1.79412343438e-06" name="Cr54" />
<nuclide ao="8.6698303856e-06" name="Fe54" />
<nuclide ao="0.000136097795928" name="Fe56" />
<nuclide ao="3.14309163167e-06" name="Fe57" />
<nuclide ao="4.18287796192e-07" name="Fe58" />
<nuclide ao="0.0218275801553" name="Zr90" />
<nuclide ao="0.00476006704262" name="Zr91" />
<nuclide ao="0.00727586005178" name="Zr92" />
<nuclide ao="0.00737343718367" name="Zr94" />
<nuclide ao="0.00118789551866" name="Zr96" />
<nuclide ao="4.67352132789e-06" name="Sn112" />
<nuclide ao="3.17992172826e-06" name="Sn114" />
<nuclide ao="1.63814149637e-06" name="Sn115" />
<nuclide ao="7.00546392861e-05" name="Sn116" />
<nuclide ao="3.70027255652e-05" name="Sn117" />
<nuclide ao="0.000116693491301" name="Sn118" />
<nuclide ao="4.13871630996e-05" name="Sn119" />
<nuclide ao="0.000156972499858" name="Sn120" />
<nuclide ao="2.230763273e-05" name="Sn122" />
<nuclide ao="2.78965860706e-05" name="Sn124" />
</material>
<material id="10014">
<temperature>300</temperature>
<density units="sum" />
<nuclide ao="0.0494673507446" name="H1" />
<nuclide ao="7.70524521986e-06" name="H2" />
<nuclide ao="7.97314977816e-06" name="B10" />
<nuclide ao="3.22546493044e-05" name="B11" />
<nuclide ao="0.0247281524718" name="O16" />
<nuclide ao="9.37552311007e-06" name="O17" />
<sab name="c_H_in_H2O" />
</material>
<material id="10015">
<temperature>300</temperature>
<density units="sum" />
<nuclide ao="2.92646478955e-06" name="U234" />
<nuclide ao="0.000342093201849" name="U235" />
<nuclide ao="5.58330430985e-06" name="U236" />
<nuclide ao="4.4860013756e-08" name="U237" />
<nuclide ao="0.0225998401942" name="U238" />
<nuclide ao="1.67593577157e-08" name="U239" />
<nuclide ao="5.10455742729e-12" name="U240" />
<nuclide ao="2.45330527464e-24" name="Np234" />
<nuclide ao="6.27669764196e-17" name="Np235" />
<nuclide ao="2.93822288698e-14" name="Np236" />
<nuclide ao="9.72443628265e-08" name="Np237" />
<nuclide ao="3.59151278099e-10" name="Np238" />
<nuclide ao="2.42672236356e-06" name="Np239" />
<nuclide ao="1.61562842872e-19" name="Pu236" />
<nuclide ao="1.67889180568e-14" name="Pu237" />
<nuclide ao="1.36231860171e-09" name="Pu238" />
<nuclide ao="1.91521644814e-05" name="Pu239" />
<nuclide ao="6.12956065082e-07" name="Pu240" />
<nuclide ao="5.2199265716e-08" name="Pu241" />
<nuclide ao="7.25971603175e-10" name="Pu242" />
<nuclide ao="2.28173101519e-13" name="Pu243" />
<nuclide ao="7.48162936285e-16" name="Pu244" />
<nuclide ao="3.33657841718e-39" name="Li6" />
<nuclide ao="1.58687690166e-30" name="Li7" />
<nuclide ao="1.56973078328e-29" name="B10" />
<nuclide ao="6.53029990405e-20" name="B11" />
<nuclide ao="6.76309478137e-15" name="N14" />
<nuclide ao="4.29750620166e-14" name="N15" />
<nuclide ao="0.0459903505854" name="O16" />
<nuclide ao="1.74396991871e-05" name="O17" />
<nuclide ao="7.76255588604e-119" name="S33" />
<nuclide ao="6.95240109507e-102" name="Cl35" />
<nuclide ao="1.54558227956e-114" name="S36" />
<nuclide ao="7.97988255929e-113" name="Ar36" />
<nuclide ao="1.05943800988e-99" name="Cl37" />
<nuclide ao="8.092276746e-93" name="Ar38" />
<nuclide ao="1.54925746209e-98" name="K39" />
<nuclide ao="1.13018880343e-91" name="Ar40" />
<nuclide ao="1.24450419477e-94" name="K40" />
<nuclide ao="2.46154453391e-106" name="Ca40" />
<nuclide ao="4.99380790875e-85" name="K41" />
<nuclide ao="2.69096671977e-86" name="Ca42" />
<nuclide ao="2.04935234814e-84" name="Ca43" />
<nuclide ao="5.86082634193e-76" name="Ca44" />
<nuclide ao="3.51777317829e-78" name="Sc45" />
<nuclide ao="3.3229675563e-75" name="Ca46" />
<nuclide ao="2.29977815684e-77" name="Ti46" />
<nuclide ao="2.68509994405e-69" name="Ti47" />
<nuclide ao="1.77753913948e-67" name="Ti48" />
<nuclide ao="8.05010452735e-67" name="Ti49" />
<nuclide ao="7.41433467824e-59" name="Ti50" />
<nuclide ao="1.47769846618e-66" name="V50" />
<nuclide ao="1.08405843777e-86" name="Cr50" />
<nuclide ao="3.5625345089e-58" name="V51" />
<nuclide ao="9.0794481129e-59" name="Cr52" />
<nuclide ao="2.32322641204e-51" name="Cr53" />
<nuclide ao="1.84676336912e-49" name="Cr54" />
<nuclide ao="3.59070585282e-54" name="Fe54" />
<nuclide ao="3.21612960797e-44" name="Mn55" />
<nuclide ao="1.20158719285e-43" name="Fe56" />
<nuclide ao="4.04601800701e-42" name="Fe57" />
<nuclide ao="8.15996548907e-36" name="Fe58" />
<nuclide ao="1.54007158722e-43" name="Co58" />
<nuclide ao="1.84676066297e-51" name="Ni58" />
<nuclide ao="1.52391142746e-35" name="Co59" />
<nuclide ao="1.53934445898e-43" name="Ni59" />
<nuclide ao="4.39068747463e-35" name="Ni60" />
<nuclide ao="4.39740720104e-29" name="Ni61" />
<nuclide ao="4.47106102555e-26" name="Ni62" />
<nuclide ao="4.68979242688e-25" name="Cu63" />
<nuclide ao="3.96684503967e-20" name="Ni64" />
<nuclide ao="5.51448239795e-29" name="Zn64" />
<nuclide ao="1.16484968718e-23" name="Cu65" />
<nuclide ao="2.81918019142e-22" name="Zn65" />
<nuclide ao="1.33210655514e-14" name="Zn66" />
<nuclide ao="8.56438124459e-14" name="Zn67" />
<nuclide ao="1.56857685926e-16" name="Zn68" />
<nuclide ao="4.85739646288e-13" name="Ga69" />
<nuclide ao="1.37392885679e-12" name="Zn70" />
<nuclide ao="3.25193501267e-16" name="Ge70" />
<nuclide ao="5.38055089065e-12" name="Ga71" />
<nuclide ao="8.28103251785e-12" name="Ge72" />
<nuclide ao="3.23954615169e-11" name="Ge73" />
<nuclide ao="1.03619341576e-10" name="Ge74" />
<nuclide ao="7.96916753088e-17" name="As74" />
<nuclide ao="1.2352453511e-17" name="Se74" />
<nuclide ao="3.1626801741e-10" name="As75" />
<nuclide ao="9.34126883389e-10" name="Ge76" />
<nuclide ao="5.08603126909e-13" name="Se76" />
<nuclide ao="2.26586227815e-09" name="Se77" />
<nuclide ao="6.23301697368e-09" name="Se78" />
<nuclide ao="1.3161721612e-08" name="Se79" />
<nuclide ao="3.14057938686e-15" name="Br79" />
<nuclide ao="3.73441121673e-08" name="Se80" />
<nuclide ao="6.98755057867e-14" name="Kr80" />
<nuclide ao="9.39903533436e-08" name="Br81" />
<nuclide ao="1.23467448441e-07" name="Se82" />
<nuclide ao="1.36421019617e-10" name="Kr82" />
<nuclide ao="1.86086471825e-07" name="Kr83" />
<nuclide ao="2.85643716296e-07" name="Kr84" />
<nuclide ao="3.20389272957e-17" name="Sr84" />
<nuclide ao="3.68409396823e-07" name="Kr85" />
<nuclide ao="7.3394176109e-09" name="Rb85" />
<nuclide ao="4.08217114533e-07" name="Kr86" />
<nuclide ao="4.73955957639e-12" name="Rb86" />
<nuclide ao="2.66313356331e-12" name="Sr86" />
<nuclide ao="5.2975042106e-07" name="Rb87" />
<nuclide ao="1.30678290809e-12" name="Sr87" />
<nuclide ao="1.01279299255e-06" name="Sr88" />
<nuclide ao="1.12190269951e-06" name="Sr89" />
<nuclide ao="2.71539686239e-07" name="Y89" />
<nuclide ao="1.50680051445e-06" name="Sr90" />
<nuclide ao="3.43880269179e-10" name="Y90" />
<nuclide ao="1.31237586182e-09" name="Zr90" />
<nuclide ao="1.37624397427e-06" name="Y91" />
<nuclide ao="2.79562047703e-07" name="Zr91" />
<nuclide ao="1.72740157381e-06" name="Zr92" />
<nuclide ao="2.17297655317e-29" name="Mo92" />
<nuclide ao="1.84175235429e-06" name="Zr93" />
<nuclide ao="4.15645403063e-14" name="Nb93" />
<nuclide ao="1.96196272967e-06" name="Zr94" />
<nuclide ao="3.53607247354e-13" name="Nb94" />
<nuclide ao="1.30767866816e-14" name="Mo94" />
<nuclide ao="1.64677967372e-06" name="Zr95" />
<nuclide ao="2.49585751652e-07" name="Nb95" />
<nuclide ao="5.81607806086e-08" name="Mo95" />
<nuclide ao="1.9040968337e-06" name="Zr96" />
<nuclide ao="7.19860476219e-10" name="Mo96" />
<nuclide ao="1.30777078645e-06" name="Mo97" />
<nuclide ao="1.80518533765e-06" name="Mo98" />
<nuclide ao="2.25393937542e-19" name="Ru98" />
<nuclide ao="2.17183365518e-07" name="Mo99" />
<nuclide ao="1.56068041817e-06" name="Tc99" />
<nuclide ao="2.0495870022e-13" name="Ru99" />
<nuclide ao="2.15048203847e-06" name="Mo100" />
<nuclide ao="7.54208363703e-09" name="Ru100" />
<nuclide ao="1.64067100694e-06" name="Ru101" />
<nuclide ao="1.41959917896e-06" name="Ru102" />
<nuclide ao="1.65645688721e-16" name="Pd102" />
<nuclide ao="8.28302115183e-07" name="Ru103" />
<nuclide ao="2.50928581819e-07" name="Rh103" />
<nuclide ao="7.63388957972e-07" name="Ru104" />
<nuclide ao="1.78487006457e-13" name="Pd104" />
<nuclide ao="4.27133693269e-09" name="Ru105" />
<nuclide ao="2.90266770652e-08" name="Rh105" />
<nuclide ao="3.63193666704e-07" name="Pd105" />
<nuclide ao="3.03115170312e-07" name="Ru106" />
<nuclide ao="6.47085862708e-08" name="Pd106" />
<nuclide ao="1.96321221839e-07" name="Pd107" />
<nuclide ao="8.63895693074e-16" name="Ag107" />
<nuclide ao="1.32214353841e-07" name="Pd108" />
<nuclide ao="1.20782261886e-14" name="Cd108" />
<nuclide ao="7.79748624648e-08" name="Ag109" />
<nuclide ao="4.68442928785e-08" name="Pd110" />
<nuclide ao="1.74146867266e-09" name="Cd110" />
<nuclide ao="7.58188335598e-09" name="Ag111" />
<nuclide ao="1.48191935748e-08" name="Cd111" />
<nuclide ao="9.92788798525e-09" name="Cd112" />
<nuclide ao="4.66251219133e-26" name="Sn112" />
<nuclide ao="2.0050526627e-09" name="Cd113" />
<nuclide ao="7.32449695913e-19" name="In113" />
<nuclide ao="4.08557527841e-23" name="Sn113" />
<nuclide ao="1.21308896349e-08" name="Cd114" />
<nuclide ao="9.48002342573e-16" name="Sn114" />
<nuclide ao="5.91187699117e-09" name="In115" />
<nuclide ao="2.77969317042e-17" name="Sn115" />
<nuclide ao="6.62700920632e-09" name="Cd116" />
<nuclide ao="3.20370125545e-10" name="Sn116" />
<nuclide ao="4.61504347035e-09" name="Sn117" />
<nuclide ao="4.7791964991e-09" name="Sn118" />
<nuclide ao="5.45802121411e-09" name="Sn119" />
<nuclide ao="4.69157936157e-09" name="Sn120" />
<nuclide ao="3.42499379543e-20" name="Te120" />
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</materials>

13
pin-cell/depleted/settings.xml
View file

@ -1,13 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<settings>
<run_mode>eigenvalue</run_mode>
<particles>10000</particles>
<batches>20</batches>
<inactive>10</inactive>
<source strength="1.0">
<space type="box">
<parameters>-0.62992 -0.62992 -10.0 0.62992 0.62992 10.0</parameters>
</space>
</source>
<seed>4553900360086534478</seed>
</settings>

11
pin-cell/depleted/tallies.xml
View file

@ -1,11 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<tallies>
<filter id="1" type="material">
<bins>10015</bins>
</filter>
<tally id="1">
<filters>1</filters>
<nuclides>Ge70 Se79 Np237 Se77 Gd156 U239 Ti50 O17 Nb95 Cd114 I135 Li7 Ca43 V51 Gd158 U236 Ca40 Se74 Br81 Zr96 Mo97 Pu239 Ba134 Ca42 Sb121 Nb93 Pu243 Np238 Se78 S33 Sn119 Sb125 Zr93 Xe131 Pd108 Pr143 Dy163 Ho165 Gd157 Te123 Zn68 Sn120 Te128 Ca46 Zn67 Sr90 Cd112 Fe57 Te122 Ni58 Pr142 Pu237 Cr52 Mn55 Cu65 Zr94 Nd142 Nd143 Gd153 H2 Zn65 Cs133 Cl35 Sc45 Ru104 Te130 Se82 Zr90 Cr53 Zr95 Pd105 Sn126 Xe130 Ar40 Ar36 Se80 Cu63 Pd102 Te125 Kr86 Sm149 Eu152 Cr50 Ce141 Ga71 Pm149 Eu155 Rh105 Ru103 Sr88 Pd104 K41 Zn64 Sn118 U235 I129 Ba135 Cr54 Ni61 Ce144 Nd150 Mo99 N14 Ru102 Ba138 Ru105 Ba136 Se76 Sn117 Tb159 Dy162 La138 Ge74 Np239 Ba132 U240 Xe129 N15 Nd147 Ce136 Tm168 Kr82 Er168 Xe132 Tm169 Zn70 Ti49 Ce140 Fe58 Zr92 Rb86 Tc99 Ce142 Tm170 Co58 Ba137 Rb85 In115 Ag109 Xe126 Gd155 Ga69 V50 Gd154 Sm150 Gd160 Pu244 Cl37 In113 Kr85 Sr89 Rh103 Pd106 Ge73 Eu154 Ru98 Er162 He3 Te120 Dy156 Dy158 Y90 Ce138 U234 Ni64 Cs135 H1 Pu241 Er167 La140 Mo92 K39 Pu236 Te126 Sm151 Cd111 Zr91 Ba140 Cd116 Ba130 Xe123 Xe133 Pr141 Ge72 Sn123 Xe136 Er170 Gd152 Nd144 Cs136 Cs134 Dy161 Cd110 Kr80 Ag111 He4 Ru100 I131 Li6 Ru106 Sm154 Tb160 Sr87 Sm147 Eu151 Nb94 Sm152 Pd110 Ar38 Eu156 Ti47 Pm147 Np234 B10 Ce139 Ni60 I130 Sr86 Sn125 Sn112 Sr84 O16 Np235 Fe54 Y91 Xe134 Eu153 Sn122 Cs137 Np236 Pu242 Ti48 Te124 Ni59 U238 U237 Fe56 Cd108 Xe128 Eu157 Mo94 Pu238 Mo100 Pu240 Xe124 Er164 Mo96 Ru99 Sb124 Sn116 Cd113 Kr84 Zn66 B11 Pd107 Co59 K40 Ag107 Ce143 Nd148 Sm148 La139 Ca44 Sm153 Dy164 Ge76 Ti46 Sn124 I127 Ba133 Sb123 As74 Sn115 As75 Nd146 Er166 Nd145 Y89 Dy160 Sn114 Ru101 Br79 Te132 Kr83 Rb87 Xe135 Pm148 Mo95 S36 Sn113 Sb126 Ni62 Mo98 Pm151</nuclides>
<scores>(n,p) (n,a) (n,gamma) fission (n,2n) (n,3n) (n,4n)</scores>
</tally>
</tallies>

15
pin-cell/fresh/geometry.xml
View file

@ -1,15 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="10053" material="10008" name="Fuel rod active region - 1.6% enr radial 0: Fuel 1.6%" region="-10034" universe="10008" />
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<surface boundary="reflective" coeffs="0.62992" id="10152" type="y-plane" />
</geometry>

267
pin-cell/fresh/materials.xml
View file

@ -1,267 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<materials>
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<nuclide name="Fe58" wo="0.0028226730001262813" />
</material>
<material id="10008" name="Fuel 1.6%">
<temperature>300</temperature>
<density units="g/cc" value="10.31341" />
<nuclide ao="1.9992419999999993" name="O16" />
<nuclide ao="0.0007579999999999998" name="O17" />
<nuclide ao="0.00013098435147670763" name="U234" />
<nuclide ao="0.01630317699531038" name="U235" />
<nuclide ao="0.9835658386532129" name="U238" />
</material>
<material id="10009" name="Fuel 2.4%">
<temperature>300</temperature>
<density units="g/cc" value="10.29748" />
<nuclide ao="1.9992420000000004" name="O16" />
<nuclide ao="0.0007580000000000002" name="O17" />
<nuclide ao="0.00019522327124383906" name="U234" />
<nuclide ao="0.024298776982208985" name="U235" />
<nuclide ao="0.9755059997465472" name="U238" />
</material>
<material id="10010" name="Fuel 3.1%">
<temperature>300</temperature>
<density units="g/cc" value="10.30166" />
<nuclide ao="1.9992420000000022" name="O16" />
<nuclide ao="0.0007580000000000009" name="O17" />
<nuclide ao="0.0002523276152188021" name="U234" />
<nuclide ao="0.03140636057161555" name="U235" />
<nuclide ao="0.9683413118131656" name="U238" />
</material>
<material id="10011" name="Fuel 3.2%">
<temperature>300</temperature>
<density units="g/cc" value="10.34115" />
<nuclide ao="1.9992419999999995" name="O16" />
<nuclide ao="0.0007579999999999998" name="O17" />
<nuclide ao="0.00025991006602092703" name="U234" />
<nuclide ao="0.03235012244921097" name="U235" />
<nuclide ao="0.9673899674847681" name="U238" />
</material>
<material id="10012" name="Fuel 3.4%">
<temperature>300</temperature>
<density units="g/cc" value="10.35917" />
<nuclide ao="1.999241999999999" name="O16" />
<nuclide ao="0.0007579999999999997" name="O17" />
<nuclide ao="0.0002770142199967032" name="U234" />
<nuclide ao="0.03447901835531245" name="U235" />
<nuclide ao="0.9652439674246908" name="U238" />
</material>
<material id="10013" name="Borosilicate Glass">
<temperature>300</temperature>
<density units="g/cc" value="2.26" />
<nuclide ao="0.013479369482225239" name="B10" />
<nuclide ao="0.054735873774104264" name="B11" />
<nuclide ao="0.6509787013744828" name="O16" />
<nuclide ao="0.00024681447050525047" name="O17" />
<nuclide ao="0.23640592474731761" name="Si28" />
<nuclide ao="0.01200401834354893" name="Si29" />
<nuclide ao="0.007913102535354443" name="Si30" />
<nuclide ao="0.024236195272461444" name="Al27" />
</material>
<material id="10014" name="Borated Water">
<temperature>300</temperature>
<density units="g/cc" value="0.7405820675158279" />
<nuclide ao="0.00032178659941803253" name="B10" />
<nuclide ao="0.0013017583017829388" name="B11" />
<nuclide ao="1.996441935899364" name="H1" />
<nuclide ao="0.0003109742982341739" name="H2" />
<nuclide ao="0.9979980704223166" name="O16" />
<nuclide ao="0.0003783846764824448" name="O17" />
<sab name="c_H_in_H2O" />
</material>
<material id="10015" name="Water SPN">
<density units="g/cc" value="0.9810025319057221" />
<nuclide ao="0.00032178659941803253" name="B10" />
<nuclide ao="0.0013017583017829388" name="B11" />
<nuclide ao="1.996441935899364" name="H1" />
<nuclide ao="0.0003109742982341739" name="H2" />
<nuclide ao="0.9979980704223166" name="O16" />
<nuclide ao="0.0003783846764824448" name="O17" />
<sab name="c_H_in_H2O" />
</material>
<material id="10016" name="SS SPN">
<temperature>300</temperature>
<density units="g/cc" value="3.6838480704877297" />
<nuclide name="Si28" wo="0.005512411036974801" />
<nuclide name="Si29" wo="0.0002899050175706448" />
<nuclide name="Si30" wo="0.00019768394545455493" />
<nuclide name="Cr50" wo="0.007930004298910168" />
<nuclide name="Cr52" wo="0.15902878860895933" />
<nuclide name="Cr53" wo="0.018379815074864116" />
<nuclide name="Cr54" wo="0.004661392017266364" />
<nuclide name="Mn55" wo="0.02" />
<nuclide name="Fe54" wo="0.03861561826636726" />
<nuclide name="Fe56" wo="0.6286064568062312" />
<nuclide name="Fe57" wo="0.014776921339264018" />
<nuclide name="Fe58" wo="0.002001003588137652" />
<nuclide name="Ni58" wo="0.06719770531879568" />
<nuclide name="Ni60" wo="0.02677596289274688" />
<nuclide name="Ni61" wo="0.0011833590846680462" />
<nuclide name="Ni62" wo="0.0038348222920813694" />
<nuclide name="Ni64" wo="0.0010081504117080411" />
</material>
</materials>

23
pin-cell/fresh/plots.xml
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<?xml version='1.0' encoding='utf-8'?>
<plots>
<plot basis="xy" color_by="material" filename="fuel-pin" id="10020" type="slice">
<origin>0.0 0.0 inf</origin>
<width>1.25984 1.25984</width>
<pixels>1000 1000</pixels>
<color id="10000" rgb="255 255 255" />
<color id="10001" rgb="0 0 0" />
<color id="10002" rgb="255 0 0" />
<color id="10003" rgb="200 50 50" />
<color id="10004" rgb="255 218 185" />
<color id="10005" rgb="101 101 101" />
<color id="10006" rgb="111 111 111" />
<color id="10007" rgb="50 50 50" />
<color id="10008" rgb="142 35 35" />
<color id="10009" rgb="255 215 0" />
<color id="10010" rgb="0 0 128" />
<color id="10013" rgb="0 255 0" />
<color id="10014" rgb="198 226 255" />
<color id="10015" rgb="176 196 222" />
<color id="10016" rgb="112 128 144" />
</plot>
</plots>

15
pin-cell/fresh/settings.xml
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<?xml version='1.0' encoding='utf-8'?>
<settings>
<run_mode>eigenvalue</run_mode>
<particles>10000</particles>
<batches>10</batches>
<inactive>5</inactive>
<source strength="1.0">
<space type="fission">
<parameters>-0.62992 -0.62992 -10.0 0.62992 0.62992 10.0</parameters>
</space>
</source>
<output>
<tallies>false</tallies>
</output>
</settings>

11
pin-cell/fresh/tallies.xml
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<?xml version='1.0' encoding='utf-8'?>
<tallies>
<filter id="1" type="material">
<bins>10008</bins>
</filter>
<tally id="10000" name="depletion tally">
<filters>1</filters>
<nuclides>O16 O17 U234 U235 U238</nuclides>
<scores>(n,p) (n,a) (n,gamma) fission (n,2n) (n,3n) (n,4n)</scores>
</tally>
</tallies>

75
pin-cell/geometry.py
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"""Creates a 2D fuel pin cell with reflective BCs."""
import numpy as np
import openmc
from beavrs.builder import BEAVRS
def find_pin(pin_name, wrap_geometry=True):
"""Find a fuel pin with some string name in the BEAVRS OpenMC model.
This method extracts the pin cell and wraps it in an OpenMC Geometry.
The returned geometry has reflective boundary conditions along the x and y
boundaries. The z-axis left unbounded.
Parameters
----------
pin_name : str
The name of the fuel pin universe
wrap_geometry : bool
If false, the pin cell Universe is returned. If true, the pin cell
Universe is wrapped in an OpenMC Geometry and returned (default).
Returns
-------
openmc.Universe
The OpenMC Universe or Geometry for this fuel pin or None if not found
"""
# Get all OpenMC Universes
all_univ = beavrs.main_universe.get_all_universes()
# Iterate over all Universes
fuel_pin = None
for univ_id, univ in all_univ.items():
if univ._name == pin_name:
fuel_pin = univ
# Wrap pin cell Universe in a Geometry if requested by the user
if wrap_geometry:
# Make reflective boundaries
pin_pitch = 0.62992
min_x = openmc.XPlane(x0=-pin_pitch, boundary_type='reflective')
max_x = openmc.XPlane(x0=pin_pitch, boundary_type='reflective')
min_y = openmc.YPlane(y0=-pin_pitch, boundary_type='reflective')
max_y = openmc.YPlane(y0=pin_pitch, boundary_type='reflective')
# Create a root Cell
root_cell = openmc.Cell(name='root cell')
root_cell.fill = fuel_pin
# Add boundaries to the root Cell
root_cell.region = +min_x & -max_x & +min_y & -max_y
# Create a root Universe
root_univ = openmc.Universe(universe_id=0, name='root universe')
root_univ.add_cell(root_cell)
# Create a Geometry
fuel_pin = openmc.Geometry(root_univ)
return fuel_pin
# User-specified enrichment of 1.6, 2.4 or 3.1 percent
enrichment = 1.6
# Instantiate a BEAVRS object
beavrs = BEAVRS()
# Extract fuel pin of interest from BEAVRS model
pin_name = 'Fuel rod active region - {}% enr'.format(enrichment)
openmc_geometry = find_pin(pin_name)

85
profile.py
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"""A parametric study over MPI procs and OMP threads for a benchmark."""
import os
import glob
import numpy as np
import openmc
# Query user for benchmark to profile (i.e., '2x2-periodic')
benchmark = input('Benchmark: ')
os.chdir(benchmark)
# Query user for hardware platform to consider
platform = os.environ['HOSTNAME']
# Default MPI+OMP runtime parameters for Phi/Haswell
if 'thing' in platform:
platform = 'haswell'
mpi_procs = [1, 2, 4, 8, 36, 72]
omp_threads = [72, 36, 18, 9, 2, 1]
else:
platform = 'phi'
mpi_procs = [4, 4, 4, 64, 64, 64]
omp_threads = [16, 32, 64, 1, 2, 4]
# Allocate arrays for timing data for inactive and active cycles
times = np.zeros((len(mpi_procs),2), dtype=np.int)
# Instantiate a Summary object to retrieve the geometry
su = openmc.Summary('summary.h5')
for i, xs in enumerate(['ace', 'multipole']):
# Construct uniform initial source distribution over fissionable zones
lower_left = su.opencg_geometry.bounds[:2] + [-10.]
upper_right = su.opencg_geometry.bounds[3:5] + [10.]
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
source.space.only_fissionable = True
settings_file = openmc.Settings()
settings_file.batches = 10
settings_file.inactive = 5
settings_file.particles = 10000
settings_file.ptables = True
settings_file.output = {'tallies': False}
settings_file.source = source
settings_file.sourcepoint_write = False
if xs == 'multipole':
settings_file.temperature = {'multipole': True, 'tolerance': 1000}
settings_file.export_to_xml()
for j, (procs, threads) in enumerate(zip(mpi_procs, omp_threads)):
# Run OpenMC - works for both flat and cache memory modes
#openmc.run(threads=threads, mpi_procs=procs,
# mpi_exec='HYDRA_TOPO_DEBUG=1 mpiexec -bind-to numa')
# Run OpenMC - works for both flat mode with MCDRAM only
#openmc.run(threads=threads, mpi_procs=procs,
# mpi_exec='HYDRA_TOPO_DEBUG=1 mpiexec -bind-to core:16 numactl --preferred 4,5,6,7')
# Run OpenMC - works for both flat and cache memory modes
openmc.run(threads=threads, mpi_procs=procs,
mpi_exec='HYDRA_TOPO_DEBUG=1 mpiexec -bind-to core:16')
# Glob the names of all statepoints in the directory
sp_filenames = glob.glob('statepoint.*.h5')
# Load the final statepoint
sp = openmc.StatePoint(sp_filenames[0])
# Extract cumulative time spent in (in)active cycles in seconds
inactive = sp.runtime['inactive batches']
active = sp.runtime['active batches']
# Convert times to neutrons / second
times[j,0] = int((sp.n_inactive * sp.n_particles) / inactive)
times[j,1] = int((sp.n_realizations * sp.n_particles) / active)
print('inactive time (n / sec): {}'.format(times[j,0]))
print('active time (n / sec): {}'.format(times[j,1]))
# Save timing data to CSV files
np.savetxt('{}-{}.csv'.format(platform, xs), times, delimiter=',', fmt='%d')

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smr/build-assembly-long.py Normal file
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#!/usr/bin/env python3
import argparse
from math import pi, isclose
from pathlib import Path
import numpy as np
from tqdm import tqdm
import openmc
from smr.materials import materials, clone
from smr.surfaces import surfs, lattice_pitch, pin_pitch, bottom_fuel_stack, \
top_active_core, pellet_OR, active_fuel_length
from smr.pins import pin_universes, make_stack
# Define command-line options
parser = argparse.ArgumentParser()
parser.add_argument('--multipole', action='store_true',
help='Use multipole cross sections')
parser.add_argument('--no-multipole', dest='multipole', action='store_false',
help='Do not use multipole cross sections')
parser.add_argument('--clone', action='store_true',
help='Clone materials for each cell instance')
parser.add_argument('--no-clone', dest='clone', action='store_false',
help='Do not clone materials for each cell instance')
parser.add_argument('-a', '--axial', type=int, default=100,
help='Number of axial subdivisions in fuel')
parser.add_argument('-d', '--depleted', action='store_true',
help='Whether UO2 compositions should represent depleted fuel')
parser.add_argument('-o', '--output-dir', type=Path, default=None)
parser.set_defaults(clone=False, multipole=True)
args = parser.parse_args()
# Make directory for inputs
if args.output_dir is None:
if args.depleted:
directory = Path('assembly-long-depleted')
else:
directory = Path('assembly-long-fresh')
else:
directory = args.output_dir
directory.mkdir(exist_ok=True)
rings = [0.1*pin_pitch, 0.2*pin_pitch]
assembly_long_surfs = [
surfs['bottom FR'],
surfs['bot active core'],
surfs['top active core'],
surfs['top pin plenum'],
surfs['top FR'],
surfs['bot upper nozzle'],
surfs['top upper nozzle']
]
univs = pin_universes(rings, args.axial, args.depleted)
fuel_univ = make_stack(
'Fuel (3.1%) stack no grid',
surfaces=assembly_long_surfs,
universes=[
univs['water pin'],
univs['end plug'],
univs['Fuel pin (3.1%) no grid'],
univs['pin plenum'],
univs['end plug'],
univs['water pin']
]
)
# Define the NumPy array indices for assembly locations where there
# may be CR guide tubes, instrument tubes and burnable absorbers
nonfuel_y = np.array([2,2,2,3,3,5,5,5,5,5,8,8,8,8,8,11,11,11,11,11,13,13,14,14,14])
nonfuel_x = np.array([5,8,11,3,13,2,5,8,11,14,2,5,8,11,14,2,5,8,11,14,3,13,5,8,11])
universes = np.full((17,17), fuel_univ)
universes[nonfuel_y, nonfuel_x] = univs['GT empty']
# Instantiate the lattice
lattice = openmc.RectLattice(name='Pin lattice')
lattice.lower_left = (-17.*pin_pitch/2., -17.*pin_pitch/2.)
lattice.pitch = (pin_pitch, pin_pitch)
lattice.universes = universes
# Add lattice to bounding cell
root_universe = openmc.Universe(name='Root universe')
cell = openmc.Cell(name='Lattice cell')
cell.fill = lattice
z_bounds = +surfs['bottom FR'] & -surfs['top FR']
cell.region = surfs['lat grid box inner'] & z_bounds
root_universe.add_cell(cell)
# Apply reflective boundaries on sides and vacuum on bottom/top
surfs['bottom FR'].boundary_type = 'vacuum'
surfs['top FR'].boundary_type = 'vacuum'
for halfspace in surfs['lat grid box inner']:
halfspace.surface.boundary_type = 'reflective'
# Define geometry with a single assembly
geometry = openmc.Geometry(root_universe)
h = active_fuel_length / args.axial
fuel_mats = {}
# Count the number of instances for each cell and material
if args.clone:
geometry.determine_paths(instances_only=True)
for cell in tqdm(geometry.get_all_material_cells().values(),
desc='Differentiating materials / assigning volume'):
if cell.fill in materials:
# Determine if this material is fuel
is_fuel = 'UO2 Fuel' in cell.fill.name
# Fill cell with list of "differentiated" materials if requested
if args.clone:
cell.fill = [clone(cell.fill) for i in range(cell.num_instances)]
# Determine volume of each fuel material
if is_fuel:
upper_right = cell.region.bounding_box[1]
if isclose(upper_right[0], rings[0]):
ri, ro = 0.0, rings[0]
elif isclose(upper_right[0], rings[1]):
ri, ro = rings[0], rings[1]
else:
ri, ro = rings[1], pellet_OR
if args.clone:
for mat in cell.fill:
mat.volume = pi * (ro*ro - ri*ri) * h
else:
# In non-clone mode, we still need to create a copy of the
# material for each ring since they get different volumes
if ri not in fuel_mats:
cell.fill = cell.fill.clone()
cell.fill.volume = pi * (ro*ro - ri*ri) * h
fuel_mats[ri] = cell.fill
else:
cell.fill = fuel_mats[ri]
else:
if args.clone:
for mat in cell.fill:
mat.volume = 1.0
else:
cell.fill.volume = 1.0
#### Create OpenMC "materials.xml" file
print('Getting materials...')
all_materials = geometry.get_all_materials()
print('Creating materials collection...')
materials = openmc.Materials(all_materials.values())
print('Exporting materials to XML...')
materials.export_to_xml(str(directory / 'materials.xml'))
#### Create OpenMC "geometry.xml" file
geometry.export_to_xml(str(directory / 'geometry.xml'))
#### Create OpenMC "settings.xml" file
# Construct uniform initial source distribution over fissionable zones
lower_left = (-lattice_pitch/2, -lattice_pitch/2, bottom_fuel_stack)
upper_right = (lattice_pitch/2, lattice_pitch/2, top_active_core)
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
source.space.only_fissionable = True
settings = openmc.Settings()
settings.batches = 200
settings.inactive = 100
settings.particles = 10000
settings.output = {'tallies': False, 'summary': False}
settings.source = source
settings.sourcepoint = {'write': False}
settings.temperature = {
'default': 531.5,
'method': 'interpolation',
'range': (500.0, 1300.0)
}
if args.multipole:
settings.temperature['multipole'] = True
settings.temperature['tolerance'] = 1000
settings.export_to_xml(str(directory / 'settings.xml'))

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smr/build-assembly-short.py Normal file
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#!/usr/bin/env python3
import argparse
from math import pi, isclose
from pathlib import Path
import numpy as np
from tqdm import tqdm
import openmc
from smr.materials import materials, clone
from smr.surfaces import surfs, lattice_pitch, pin_pitch, bottom_fuel_stack, \
top_active_core, pellet_OR, clad_OR, clad_IR, guide_tube_IR, guide_tube_OR
from smr.pins import pin_universes
# Define command-line options
parser = argparse.ArgumentParser()
parser.add_argument('--multipole', action='store_true',
help='Use multipole cross sections')
parser.add_argument('--no-multipole', action='store_false',
help='Do not use multipole cross sections')
parser.add_argument('-t', '--tallies', choices=('cell', 'mat'), default='mat',
help='Whether to use distribmats or distribcells for tallies')
parser.add_argument('-a', '--axial', type=int, default=10,
help='Number of axial subdivisions in fuel')
parser.add_argument('-d', '--depleted', action='store_true',
help='Whether UO2 compositions should represent depleted fuel')
parser.add_argument('-o', '--output-dir', type=Path, default=None)
parser.set_defaults(multipole=True)
args = parser.parse_args()
# Make directory for inputs
if args.output_dir is None:
if args.depleted:
directory = Path('assembly-short-depleted')
else:
directory = Path('assembly-short-fresh')
else:
directory = args.output_dir
directory.mkdir(exist_ok=True)
rings = [0.1*pin_pitch, 0.2*pin_pitch]
# Define the NumPy array indices for assembly locations where there
# may be CR guide tubes, instrument tubes and burnable absorbers
nonfuel_y = np.array([2,2,2,3,3,5,5,5,5,5,8,8,8,8,8,11,11,11,11,11,13,13,14,14,14])
nonfuel_x = np.array([5,8,11,3,13,2,5,8,11,14,2,5,8,11,14,2,5,8,11,14,3,13,5,8,11])
# NO BURNABLE ABSORBERS
pins = pin_universes(rings, args.axial, args.depleted)
gtu = pins['GT empty']
#gti = pins['GT empty instr']
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = pins['Fuel pin (3.1%) no grid']
universes[nonfuel_y, nonfuel_x] = [ gtu, gtu, gtu,
gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu,
gtu, gtu, gtu ]
# Instantiate the lattice
lattice = openmc.RectLattice(name='Pin lattice')
lattice.lower_left = (-17.*pin_pitch/2., -17.*pin_pitch/2.)
lattice.pitch = (pin_pitch, pin_pitch)
lattice.universes = universes
# Add lattice to bounding cell
root_universe = openmc.Universe(name='Root universe')
cell = openmc.Cell(name='Lattice cell')
cell.fill = lattice
z_bounds = +surfs['bot active core'] & -surfs['top active core']
cell.region = surfs['lat grid box inner'] & z_bounds
root_universe.add_cell(cell)
# Apply reflective boundaries
surfs['bot active core'].boundary_type = 'reflective'
surfs['top active core'].boundary_type = 'reflective'
for halfspace in surfs['lat grid box inner']:
halfspace.surface.boundary_type = 'reflective'
# Define geometry with a single assembly
geometry = openmc.Geometry(root_universe)
#### "Differentiate" the geometry if using distribmats
h = 10.0*pin_pitch / args.axial
if args.tallies == 'mat':
# Count the number of instances for each cell and material
geometry.determine_paths(instances_only=True)
for cell in tqdm(geometry.get_all_material_cells().values(),
desc='Differentiating materials'):
if cell.fill in materials:
# Fill cell with list of "differentiated" materials
cell.fill = [clone(cell.fill) for i in range(cell.num_instances)]
# Determine volume of each fuel material
if 'UO2 Fuel' in cell.fill[0].name:
lower_left, _ = cell.region.bounding_box
if isclose(lower_left[0], rings[0]):
ri, ro = 0.0, rings[0]
elif isclose(lower_left[0], rings[1]):
ri, ro = rings[0], rings[1]
else:
ri, ro = rings[1], pellet_OR
for mat in cell.fill:
mat.volume = pi * (ro*ro - ri*ri) * h
elif cell.fill[0].name == 'Borated Water':
for mat in cell.fill:
mat.volume = pin_pitch**2 - pi*clad_OR**2 * h
elif cell.fill[0].name == 'Helium':
for mat in cell.fill:
mat.volume = pi * (clad_IR**2 - pellet_OR**2) * h
elif cell.fill[0].name == 'M5':
for mat in cell.fill:
mat.volume = pi * (clad_OR**2 - clad_IR**2) * h
elif cell.fill[0].name == 'Zircaloy-4':
for mat in cell.fill:
mat.volume = pi * (guide_tube_OR**2 - guide_tube_IR**2) * h
#### Create OpenMC "materials.xml" file
print('Getting materials...')
all_materials = geometry.get_all_materials()
print('Creating materials collection...')
materials = openmc.Materials(all_materials.values())
print('Exporting materials to XML...')
materials.export_to_xml(str(directory / 'materials.xml'))
#### Create OpenMC "geometry.xml" file
geometry.export_to_xml(str(directory / 'geometry.xml'))
#### Create OpenMC "settings.xml" file
# Construct uniform initial source distribution over fissionable zones
lower_left = (-lattice_pitch/2, -lattice_pitch/2, bottom_fuel_stack)
upper_right = (lattice_pitch/2, lattice_pitch/2, top_active_core)
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
source.space.only_fissionable = True
settings = openmc.Settings()
settings.batches = 200
settings.inactive = 100
settings.particles = 10000
settings.output = {'tallies': False, 'summary': False}
settings.source = source
settings.sourcepoint = {'write': False}
if args.multipole:
settings.temperature = {
'multipole': True,
'tolerance': 1000,
'default': 531.5,
'method': 'interpolation',
'range': (500.0, 1300.0)
}
settings.export_to_xml(str(directory / 'settings.xml'))
#### Create OpenMC "tallies.xml" file
tallies = openmc.Tallies()
# Extract all fuel materials
materials = geometry.get_materials_by_name(name='Fuel', matching=False)
# If using distribcells, create distribcell tally needed for depletion
if args.tallies == 'cell':
# Extract all cells filled by a fuel material
fuel_cells = []
for cell in geometry.get_all_cells().values():
if cell.fill in materials:
tally = openmc.Tally(name='depletion tally')
tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
'fission', '(n,2n)', '(n,3n)', '(n,4n)']
tally.nuclides = cell.fill.get_nuclides()
tally.filters.append(openmc.DistribcellFilter([cell]))
tallies.append(tally)
# If using distribmats, create material tally needed for depletion
elif args.tallies == 'mat':
tally = openmc.Tally(name='depletion tally')
tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
'fission', '(n,2n)', '(n,3n)', '(n,4n)']
tally.nuclides = materials[0].get_nuclides()
tally.filters = [openmc.MaterialFilter(materials)]
tallies.append(tally)
tallies.export_to_xml(str(directory / 'tallies.xml'))

145
smr/build-assembly.py Normal file
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@ -0,0 +1,145 @@
#!/usr/bin/env python3
import argparse
from pathlib import Path
import numpy as np
from tqdm import tqdm
import openmc
from smr.materials import materials, clone
from smr.surfaces import surfs, lattice_pitch, bottom_fuel_stack, top_active_core, pellet_OR
from smr.assemblies import assembly_universes
from smr import inlet_temperature
# Define command-line options
parser = argparse.ArgumentParser()
parser.add_argument('--multipole', action='store_true',
help='Use multipole cross sections')
parser.add_argument('--no-multipole', dest='multipole', action='store_false',
help='Do not use multipole cross sections')
parser.add_argument('--clone', action='store_true',
help='Clone materials for each cell instance')
parser.add_argument('--no-clone', dest='clone', action='store_false',
help='Do not clone materials for each cell instance')
parser.add_argument('-t', '--tallies', choices=('cell', 'mat'), default='mat',
help='Whether to use distribmats or distribcells for tallies')
parser.add_argument('-r', '--rings', type=int, default=10,
help='Number of annular regions in fuel')
parser.add_argument('-a', '--axial', type=int, default=196,
help='Number of axial subdivisions in fuel')
parser.add_argument('-d', '--depleted', action='store_true',
help='Whether UO2 compositions should represent depleted fuel')
parser.add_argument('-o', '--output-dir', type=Path, default=None)
parser.set_defaults(clone=False, multipole=True)
args = parser.parse_args()
# Make directory for inputs
if args.output_dir is None:
if args.depleted:
directory = Path('assembly-depleted')
else:
directory = Path('assembly-fresh')
else:
directory = args.output_dir
directory.mkdir(exist_ok=True)
# Define geometry with a single assembly
if args.rings > 1:
ring_radii = np.sqrt(np.arange(1, args.rings)*pellet_OR**2 / args.rings)
else:
ring_radii = None
assembly = assembly_universes(ring_radii, args.axial, args.depleted)
lattice_sides = openmc.model.rectangular_prism(lattice_pitch, lattice_pitch,
boundary_type='reflective')
main_cell = openmc.Cell(
fill=assembly['Assembly (3.1%)'],
region=lattice_sides & +surfs['lower bound'] & -surfs['upper bound']
)
root_univ = openmc.Universe(cells=[main_cell])
geometry = openmc.Geometry(root_univ)
#### "Differentiate" the geometry if using distribmats
if args.clone:
# Count the number of instances for each cell and material
geometry.determine_paths(instances_only=True)
# Extract all cells filled by a fuel material
fuel_mats = {m for m in materials if 'UO2 Fuel' in m.name}
for cell in tqdm(geometry.get_all_material_cells().values(),
desc='Differentiating materials'):
if cell.fill in fuel_mats:
# Fill cell with list of "differentiated" materials
cell.fill = [clone(cell.fill) for i in range(cell.num_instances)]
#### Create OpenMC "materials.xml" file
print('Getting materials...')
all_materials = geometry.get_all_materials()
print('Creating materials collection...')
materials = openmc.Materials(all_materials.values())
print('Exporting materials to XML...')
materials.export_to_xml(str(directory / 'materials.xml'))
#### Create OpenMC "geometry.xml" file
geometry.export_to_xml(str(directory / 'geometry.xml'))
#### Create OpenMC "settings.xml" file
# Construct uniform initial source distribution over fissionable zones
lower_left = (-lattice_pitch/2, -lattice_pitch/2, bottom_fuel_stack)
upper_right = (lattice_pitch/2, lattice_pitch/2, top_active_core)
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
source.space.only_fissionable = True
settings = openmc.Settings()
settings.batches = 200
settings.inactive = 100
settings.particles = 10000
settings.output = {'tallies': False, 'summary': False}
settings.source = source
settings.sourcepoint = {'write': False}
settings.temperature = {
'default': inlet_temperature,
'method': 'interpolation',
'range': (300.0, 1500.0),
}
if args.multipole:
settings.temperature['multipole'] = True
settings.temperature['tolerance'] = 1000
settings.export_to_xml(str(directory / 'settings.xml'))
#### Create OpenMC "tallies.xml" file
tallies = openmc.Tallies()
# Extract all fuel materials
materials = geometry.get_materials_by_name(name='Fuel', matching=False)
# If using distribcells, create distribcell tally needed for depletion
if args.tallies == 'cell':
# Extract all cells filled by a fuel material
fuel_cells = []
for cell in geometry.get_all_cells().values():
if cell.fill in materials:
tally = openmc.Tally(name='depletion tally')
tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
'fission', '(n,2n)', '(n,3n)', '(n,4n)']
tally.nuclides = cell.fill.get_nuclides()
tally.filters.append(openmc.DistribcellFilter([cell]))
tallies.append(tally)
# If using distribmats, create material tally needed for depletion
elif args.tallies == 'mat':
tally = openmc.Tally(name='depletion tally')
tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
'fission', '(n,2n)', '(n,3n)', '(n,4n)']
tally.nuclides = materials[0].get_nuclides()
tally.filters = [openmc.MaterialFilter(materials)]
tallies.append(tally)
tallies.export_to_xml(str(directory / 'tallies.xml'))

2
smr/build-depleted.py → smr/build-core-depleted.py
View file

@ -64,7 +64,7 @@ settings.entropy_dimension = [15, 15, 1]
# MeV/second cm from CASMO
settings.power = 2.337e15 * ((17.*17.*37.) / 1.5**2) * height
settings.dt_vec = dt
settings.output_dir = 'depleted'
settings.output_dir = 'core-depleted'
op = opendeplete.OpenMCOperator(geometry, settings)

121
smr/build-core-fresh.py Normal file
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#!/usr/bin/env python3
import argparse
from math import pi
from pathlib import Path
import numpy as np
import openmc
from tqdm import tqdm
from smr.materials import materials, clone
from smr.surfaces import lattice_pitch, bottom_fuel_stack, top_active_core, \
pellet_OR, active_fuel_length
from smr.core import core_geometry
from smr import inlet_temperature
# Define command-line options
parser = argparse.ArgumentParser()
parser.add_argument('--multipole', action='store_true',
help='Use multipole cross sections')
parser.add_argument('--no-multipole', dest='multipole', action='store_false',
help='Do not use multipole cross sections')
parser.add_argument('--clone', action='store_true',
help='Clone materials for each cell instance')
parser.add_argument('--no-clone', dest='clone', action='store_false',
help='Do not clone materials for each cell instance')
parser.add_argument('-r', '--rings', type=int, default=10,
help='Number of annular regions in fuel')
parser.add_argument('-a', '--axial', type=int, default=196,
help='Number of axial subdivisions in fuel')
parser.add_argument('-d', '--depleted', action='store_true',
help='Whether UO2 compositions should represent depleted fuel')
parser.add_argument('-o', '--output-dir', type=Path, default=None)
parser.set_defaults(clone=False, multipole=True)
args = parser.parse_args()
# Make directory for inputs
if args.output_dir is None:
if args.depleted:
directory = Path('core-depleted')
else:
directory = Path('core-fresh')
else:
directory = args.output_dir
directory.mkdir(exist_ok=True)
if args.rings > 1:
ring_radii = np.sqrt(np.arange(1, args.rings)*pellet_OR**2 / args.rings)
else:
ring_radii = None
geometry = core_geometry(ring_radii, args.axial, args.depleted)
h = active_fuel_length / args.axial
fuel_mats = {}
# Count the number of instances for each cell and material
if args.clone:
geometry.determine_paths(instances_only=True)
fuel_volume = pi * pellet_OR**2 * h / args.rings
for cell in tqdm(geometry.get_all_cells().values(),
desc='Differentiating materials / assigning volume'):
if cell.fill in materials:
# Determine if this material is fuel
name = cell.fill.name
is_fuel = 'UO2 Fuel' in name
# Determine volume of each fuel material
if is_fuel:
if args.clone:
# Fill cell with list of "differentiated" materials if requested
cell.fill = [clone(cell.fill) for i in range(cell.num_instances)]
for mat in cell.fill:
mat.volume = fuel_volume
else:
r_o = cell.region.bounding_box[1][0]
if (name, r_o) not in fuel_mats:
cell.fill = cell.fill.clone()
cell.fill.volume = fuel_volume
fuel_mats[name, r_o] = cell.fill
else:
cell.fill = fuel_mats[name, r_o]
else:
cell.fill.volume = 1.0
#### Create OpenMC "materials.xml" file
all_materials = geometry.get_all_materials()
materials = openmc.Materials(all_materials.values())
materials.export_to_xml(str(directory / 'materials.xml'))
#### Create OpenMC "geometry.xml" file
geometry.export_to_xml(str(directory / 'geometry.xml'))
#### Create OpenMC "settings.xml" file
# Construct uniform initial source distribution over fissionable zones
lower_left = [-7.*lattice_pitch/2., -7.*lattice_pitch/2., bottom_fuel_stack]
upper_right = [+7.*lattice_pitch/2., +7.*lattice_pitch/2., top_active_core]
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
source.space.only_fissionable = True
settings = openmc.Settings()
settings.batches = 200
settings.inactive = 100
settings.particles = 10000
settings.output = {'tallies': False, 'summary': False}
settings.source = source
settings.sourcepoint = {'write': False}
settings.temperature = {
'default': inlet_temperature,
'method': 'interpolation',
'range': (300.0, 1500.0),
}
if args.multipole:
settings.temperature['multipole'] = True
settings.temperature['tolerance'] = 1000
settings.export_to_xml(str(directory / 'settings.xml'))

118
smr/build-core-long.py Normal file
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@ -0,0 +1,118 @@
#!/usr/bin/env python3
import argparse
from math import pi, isclose
from pathlib import Path
import openmc
from smr.materials import materials
from smr.surfaces import bottom_fuel_stack, top_active_core, \
pellet_OR, pin_pitch, clad_IR, clad_OR, active_fuel_length
from smr.core import core_geometry
from smr import inlet_temperature
import smr.surfaces
# Define command-line options
parser = argparse.ArgumentParser()
parser.add_argument('--multipole', action='store_true',
help='Use multipole cross sections')
parser.add_argument('--no-multipole', dest='multipole', action='store_false',
help='Do not use multipole cross sections')
parser.add_argument('-a', '--axial', type=int, default=100,
help='Number of axial subdivisions in fuel')
parser.add_argument('-d', '--depleted', action='store_true',
help='Whether UO2 compositions should represent depleted fuel')
parser.add_argument('-o', '--output-dir', type=Path, default=None)
parser.set_defaults(multipole=True)
args = parser.parse_args()
# Make directory for inputs
if args.output_dir is None:
if args.depleted:
directory = Path('core-long-depleted')
else:
directory = Path('core-long-fresh')
else:
directory = args.output_dir
directory.mkdir(exist_ok=True)
# Modify lattice pitch
smr.surfaces.lattice_pitch = lattice_pitch = 17*smr.surfaces.pin_pitch
ring_radii = [0.1*pin_pitch, 0.2*pin_pitch]
geometry = core_geometry(ring_radii, args.axial, args.depleted)
h = active_fuel_length / args.axial
fuel_mats = {}
for cell in geometry.get_all_cells().values():
if cell.fill in materials:
# Determine volume of each fuel material
name = cell.fill.name
if 'UO2 Fuel' in name:
upper_right = cell.region.bounding_box[1][0]
if isclose(upper_right, ring_radii[0]):
ri, ro = 0.0, ring_radii[0]
elif isclose(upper_right, ring_radii[1]):
ri, ro = ring_radii[0], ring_radii[1]
else:
ri, ro = ring_radii[1], pellet_OR
if (name, ri) not in fuel_mats:
cell.fill = cell.fill.clone()
cell.fill.volume = pi * (ro*ro - ri*ri) * h
fuel_mats[name, ri] = cell.fill
else:
cell.fill = fuel_mats[name, ri]
elif name == 'Helium':
cell.fill.volume = pi * (clad_IR**2 - pellet_OR**2) * h
elif name == 'M5':
# Clad is not subdivided
cell.fill.volume = pi * (clad_OR**2 - clad_IR**2) * active_fuel_length
else:
cell.fill.volume = 1.0
#### Create OpenMC "materials.xml" file
all_materials = geometry.get_all_materials()
materials = openmc.Materials(all_materials.values())
materials.export_to_xml(str(directory / 'materials.xml'))
#### Create OpenMC "geometry.xml" file
geometry.export_to_xml(str(directory / 'geometry.xml'))
#### Create OpenMC "settings.xml" file
# Construct uniform initial source distribution over fissionable zones
lower_left = [-7.*lattice_pitch/2., -7.*lattice_pitch/2., bottom_fuel_stack]
upper_right = [+7.*lattice_pitch/2., +7.*lattice_pitch/2., top_active_core]
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
source.space.only_fissionable = True
settings = openmc.Settings()
settings.batches = 200
settings.inactive = 100
settings.particles = 20_000_000
settings.output = {'tallies': False, 'summary': False}
settings.source = source
settings.sourcepoint = {'write': False}
settings.temperature = {
'default': inlet_temperature,
'method': 'interpolation',
'range': (300.0, 1500.0),
}
if args.multipole:
settings.temperature['multipole'] = True
settings.temperature['tolerance'] = 1000
settings.export_to_xml(str(directory / 'settings.xml'))
# Check assembly power distribution
core_lattice = geometry.get_cells_by_fill_name('Main core')[0].fill
mesh = openmc.RegularMesh.from_rect_lattice(core_lattice)
assembly_power = openmc.Tally()
assembly_power.filters = [openmc.MeshFilter(mesh)]
assembly_power.scores = ['nu-fission']
tallies = openmc.Tallies([assembly_power])
tallies.export_to_xml(directory / 'tallies.xml')

131
smr/build-core-short.py Normal file
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@ -0,0 +1,131 @@
#!/usr/bin/env python3
import argparse
from math import pi, isclose
from pathlib import Path
import openmc
from smr.materials import materials
from smr.surfaces import bottom_fuel_stack, top_active_core, \
pellet_OR, surfs, pin_pitch, clad_IR, clad_OR
import smr.surfaces
import smr.pins
from smr.core import core_geometry
from smr import inlet_temperature
# Define command-line options
parser = argparse.ArgumentParser()
parser.add_argument('--multipole', action='store_true',
help='Use multipole cross sections')
parser.add_argument('--no-multipole', action='store_false',
help='Do not use multipole cross sections')
parser.add_argument('-a', '--axial', type=int, default=3,
help='Number of axial subdivisions in fuel')
parser.add_argument('-d', '--depleted', action='store_true',
help='Whether UO2 compositions should represent depleted fuel')
parser.add_argument('-o', '--output-dir', type=Path, default=None)
parser.set_defaults(multipole=True)
args = parser.parse_args()
# Make directory for inputs
if args.output_dir is None:
if args.depleted:
directory = Path('core-short-depleted')
else:
directory = Path('core-short-fresh')
else:
directory = args.output_dir
directory.mkdir(exist_ok=True)
# Modify lattice pitch
smr.surfaces.lattice_pitch = lattice_pitch = 17*smr.surfaces.pin_pitch
# Modify fuel length
length = 3. * pin_pitch
smr.surfaces.active_fuel_length = length
smr.pins.top_active_core = length
surfs['top active core'].z0 = length
# Change top and bottom of model to contain only fuel
surfs['lower bound'].z0 = 0.0
surfs['lower bound'].boundary_type = 'reflective'
surfs['upper bound'].z0 = length
surfs['upper bound'].boundary_type = 'reflective'
ring_radii = [0.1*pin_pitch, 0.2*pin_pitch]
geometry = core_geometry(ring_radii, args.axial, args.depleted)
h = length / args.axial
fuel_mats = {}
for cell in geometry.get_all_cells().values():
if cell.fill in materials:
# Determine volume of each fuel material
name = cell.fill.name
if 'UO2 Fuel' in name:
upper_right = cell.region.bounding_box[1][0]
if isclose(upper_right, ring_radii[0]):
ri, ro = 0.0, ring_radii[0]
elif isclose(upper_right, ring_radii[1]):
ri, ro = ring_radii[0], ring_radii[1]
else:
ri, ro = ring_radii[1], pellet_OR
if (name, ri) not in fuel_mats:
cell.fill = cell.fill.clone()
cell.fill.volume = pi * (ro*ro - ri*ri) * h
fuel_mats[name, ri] = cell.fill
else:
cell.fill = fuel_mats[name, ri]
elif name == 'Helium':
cell.fill.volume = pi * (clad_IR**2 - pellet_OR**2) * h
elif name == 'M5':
# Clad is not subdivided
cell.fill.volume = pi * (clad_OR**2 - clad_IR**2) * length
else:
cell.fill.volume = 1.0
#### Create OpenMC "materials.xml" file
all_materials = geometry.get_all_materials()
materials = openmc.Materials(all_materials.values())
materials.export_to_xml(str(directory / 'materials.xml'))
#### Create OpenMC "geometry.xml" file
geometry.export_to_xml(str(directory / 'geometry.xml'))
#### Create OpenMC "settings.xml" file
# Construct uniform initial source distribution over fissionable zones
lower_left = [-7.*lattice_pitch/2., -7.*lattice_pitch/2., bottom_fuel_stack]
upper_right = [+7.*lattice_pitch/2., +7.*lattice_pitch/2., top_active_core]
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
source.space.only_fissionable = True
settings = openmc.Settings()
settings.batches = 200
settings.inactive = 100
settings.particles = 10000
settings.output = {'tallies': False, 'summary': False}
settings.source = source
settings.sourcepoint = {'write': False}
settings.temperature = {
'default': inlet_temperature,
'method': 'interpolation',
'range': (300.0, 1500.0),
}
if args.multipole:
settings.temperature['multipole'] = True
settings.temperature['tolerance'] = 1000
settings.export_to_xml(str(directory / 'settings.xml'))
# Check assembly power distribution
core_lattice = geometry.get_cells_by_fill_name('Main core')[0].fill
mesh = openmc.RegularMesh.from_rect_lattice(core_lattice)
assembly_power = openmc.Tally()
assembly_power.filters = [openmc.MeshFilter(mesh)]
assembly_power.scores = ['nu-fission']
tallies = openmc.Tallies([assembly_power])
tallies.export_to_xml(directory / 'tallies.xml')

163
smr/build-fresh.py
View file

@ -1,163 +0,0 @@
#!/usr/bin/env python3
import os
import shutil
import copy
import numpy as np
import openmc
from smr.materials import materials
from smr.plots import plots
from smr.surfaces import lattice_pitch, bottom_fuel_stack, top_active_core
from smr.core import geometry
#### Query the user for options
# Query the user on whether to use multipole cross sections
multipole = input('Use multipole cross sections? (y/n): ').lower()
multipole = (multipole == 'y')
# Query the user on whether to use distribmats or distribcells
# If using distribmats, the geometry must be "differentiated" with unique
# material instances for each instance of a fuel cell
distrib = input('Use distribmat or distribcells? [mat/cell]: ').lower()
if distrib not in ['cell', 'mat']:
raise InputError('Distrib type "{}" is unsupported'.format(distrib))
#### "Differentiate" the geometry if using distribmats
if distrib == 'mat':
# Count the number of instances for each cell and material
geometry.determine_paths()
# Extract all cells filled by a fuel material
fuel_cells = geometry.get_cells_by_name(
name='(1.6%) (0)', case_sensitive=True)
fuel_cells.extend(geometry.get_cells_by_name(
name='(1.6%) grid (bottom) (0)', case_sensitive=True))
fuel_cells.extend(geometry.get_cells_by_name(
name='(1.6%) grid (intermediate) (0)', case_sensitive=True))
fuel_cells.extend(geometry.get_cells_by_name(
name='(2.4%) (0)', case_sensitive=True))
fuel_cells.extend(geometry.get_cells_by_name(
name='(2.4%) grid (bottom) (0)', case_sensitive=True))
fuel_cells.extend(geometry.get_cells_by_name(
name='(2.4%) grid (intermediate) (0)', case_sensitive=True))
fuel_cells.extend(geometry.get_cells_by_name(
name='(3.1%) (0)', case_sensitive=True))
fuel_cells.extend(geometry.get_cells_by_name(
name='(3.1%) grid (bottom) (0)', case_sensitive=True))
fuel_cells.extend(geometry.get_cells_by_name(
name='(3.1%) grid (intermediate) (0)', case_sensitive=True))
# Assign distribmats for each material
for cell in fuel_cells:
new_materials = []
for i in range(cell.num_instances):
new_materials.append(cell.fill.clone())
# Fill cell with list of "differentiated" materials
cell.fill = new_materials
#### Create OpenMC "materials.xml" file
all_materials = geometry.get_all_materials()
materials = openmc.Materials(all_materials.values())
materials.export_to_xml()
#### Create OpenMC "geometry.xml" file
geometry.export_to_xml()
#### Create OpenMC "settings.xml" file
# Construct uniform initial source distribution over fissionable zones
lower_left = [-7.*lattice_pitch/2., -7.*lattice_pitch/2., bottom_fuel_stack]
upper_right = [+7.*lattice_pitch/2., +7.*lattice_pitch/2., top_active_core]
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
source.space.only_fissionable = True
settings = openmc.Settings()
settings.batches = 200
settings.inactive = 100
settings.particles = 10000
settings.output = {'tallies': False}
settings.source = source
settings.sourcepoint_write = False
if multipole:
settings.temperature = {'multipole': True, 'tolerance': 1000}
settings.export_to_xml()
#### Create OpenMC "plots.xml" file
plots.export_to_xml()
#### Create OpenMC "tallies.xml" file
tallies = openmc.Tallies()
# Extract all fuel materials
materials = geometry.get_materials_by_name(name='Fuel', matching=False)
# If using distribcells, create distribcell tally needed for depletion
if distrib == 'cell':
# Extract all cells filled by a fuel material
fuel_cells = geometry.get_cells_by_name(
name='(1.6%) (0)', case_sensitive=True)
fuel_cells.extend(geometry.get_cells_by_name(
name='(1.6%) grid (bottom) (0)', case_sensitive=True))
fuel_cells.extend(geometry.get_cells_by_name(
name='(1.6%) grid (intermediate) (0)', case_sensitive=True))
fuel_cells.extend(geometry.get_cells_by_name(
name='(2.4%) (0)', case_sensitive=True))
fuel_cells.extend(geometry.get_cells_by_name(
name='(2.4%) grid (bottom) (0)', case_sensitive=True))
fuel_cells.extend(geometry.get_cells_by_name(
name='(2.4%) grid (intermediate) (0)', case_sensitive=True))
fuel_cells.extend(geometry.get_cells_by_name(
name='(3.1%) (0)', case_sensitive=True))
fuel_cells.extend(geometry.get_cells_by_name(
name='(3.1%) grid (bottom) (0)', case_sensitive=True))
fuel_cells.extend(geometry.get_cells_by_name(
name='(3.1%) grid (intermediate) (0)', case_sensitive=True))
for cell in fuel_cells:
tally = openmc.Tally(name='depletion tally')
tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
'fission', '(n,2n)', '(n,3n)', '(n,4n)']
tally.nuclides = cell.fill.get_nuclides()
tally.filters.append(openmc.DistribcellFilter([cell.id]))
tallies.append(tally)
# If using distribmats, create material tally needed for depletion
elif distrib == 'mat':
tally = openmc.Tally(name='depletion tally')
tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
'fission', '(n,2n)', '(n,3n)', '(n,4n)']
tally.nuclides = materials[0].get_nuclides()
material_ids = [material.id for material in materials]
tally.filters.append(openmc.MaterialFilter(material_ids))
tallies.append(tally)
tallies.export_to_xml()
#### Move all XML files to 'fresh' directory
if not os.path.exists('fresh'):
os.makedirs('fresh')
shutil.move('materials.xml', 'fresh/materials.xml')
shutil.move('geometry.xml', 'fresh/geometry.xml')
shutil.move('settings.xml', 'fresh/settings.xml')
shutil.move('tallies.xml', 'fresh/tallies.xml')
shutil.move('plots.xml', 'fresh/plots.xml')

0
smr/depleted/geometry.xml → smr/core-depleted/geometry.xml
View file

0
smr/depleted/materials.xml → smr/core-depleted/materials.xml
View file

0
smr/depleted/settings.xml → smr/core-depleted/settings.xml
View file

0
smr/depleted/tallies.xml → smr/core-depleted/tallies.xml
View file

7407
smr/core-fresh/geometry.xml Normal file
View file

File diff suppressed because it is too large Load diff

4
smr/fresh/materials.xml → smr/core-fresh/materials.xml
View file

@ -172,14 +172,14 @@
<material id="9" name="Borosilicate Glass">
<temperature>300</temperature>
<density units="g/cc" value="2.26" />
<nuclide ao="0.012259454378427138" name="B10" />
<nuclide ao="0.06018253698401973" name="B11" />
<nuclide ao="0.6509787013744828" name="O16" />
<nuclide ao="0.00024681447050525047" name="O17" />
<nuclide ao="0.23640592474731761" name="Si28" />
<nuclide ao="0.01200401834354893" name="Si29" />
<nuclide ao="0.007913102535354443" name="Si30" />
<nuclide ao="0.024236195272461444" name="Al27" />
<nuclide ao="0.012259454378427138" name="B10" />
<nuclide ao="0.06018253698401973" name="B11" />
</material>
<material id="10" name="1.6% Enr. UO2 Fuel">
<temperature>300</temperature>

13
smr/fresh/plots.xml → smr/core-fresh/plots.xml
View file

@ -76,4 +76,17 @@
<color id="11" rgb="255 215 0" />
<color id="12" rgb="0 0 128" />
</plot>
<!--assembly no spacer cell-->
<plot basis="xy" color_by="cell" filename="assm_no_spacer_cell" id="5" type="slice">
<origin>0.0 0.0 90.0</origin>
<width>32.25546 32.25546</width>
<pixels>2000 2000</pixels>
<background>255 255 255</background>
</plot>
<!--z slice-->
<plot basis="xz" color_by="cell" filename="assm_xz" id="6" type="slice">
<origin>0.0 0.0 141.61599999999999</origin>
<width>32.25546 283.23199999999997</width>
<pixels>455 4000</pixels>
</plot>
</plots>

0
smr/fresh/settings.xml → smr/core-fresh/settings.xml
View file

51747
smr/core-fresh/tallies.xml Normal file
View file

File diff suppressed because it is too large Load diff

751
smr/fresh/geometry.xml
View file

@ -1,751 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="1" material="8" name="water pin" universe="1" />
<cell id="2" material="8" name="GT empty (0)" region="-5" universe="2" />
<cell id="3" material="6" name="GT empty (1)" region="5 -6" universe="2" />
<cell id="4" material="8" name="GT empty (last)" region="6" universe="2" />
<cell id="5" material="8" name="GT empty grid (bottom) (0)" region="-5" universe="3" />
<cell id="6" material="6" name="GT empty grid (bottom) (1)" region="5 -6" universe="3" />
<cell id="7" material="8" name="GT empty grid (bottom) (last)" region="6 20 -21 22 -23" universe="3" />
<cell id="8" material="3" name="GT empty grid (bottom) (grid)" region="~(20 -21 22 -23)" universe="3" />
<cell id="9" material="8" name="GT empty grid (intermediate) (0)" region="-5" universe="4" />
<cell id="10" material="6" name="GT empty grid (intermediate) (1)" region="5 -6" universe="4" />
<cell id="11" material="8" name="GT empty grid (intermediate) (last)" region="6 20 -21 22 -23" universe="4" />
<cell id="12" material="6" name="GT empty grid (intermediate) (grid)" region="~(20 -21 22 -23)" universe="4" />
<cell id="13" material="8" name="GT empty nozzle (0)" region="-5" universe="5" />
<cell id="14" material="6" name="GT empty nozzle (1)" region="5 -6" universe="5" />
<cell id="15" material="8" name="GT empty nozzle (last)" region="6" universe="5" />
<cell id="16" material="8" name="GT empty at dashpot (0)" region="-7" universe="6" />
<cell id="17" material="6" name="GT empty at dashpot (1)" region="7 -8" universe="6" />
<cell id="18" material="8" name="GT empty at dashpot (last)" region="8" universe="6" />
<cell id="19" material="8" name="GT empty at dashpot grid (bottom) (0)" region="-7" universe="7" />
<cell id="20" material="6" name="GT empty at dashpot grid (bottom) (1)" region="7 -8" universe="7" />
<cell id="21" material="8" name="GT empty at dashpot grid (bottom) (last)" region="8 20 -21 22 -23" universe="7" />
<cell id="22" material="3" name="GT empty at dashpot grid (bottom) (grid)" region="~(20 -21 22 -23)" universe="7" />
<cell id="23" material="8" name="GT empty at dashpot grid (intermediate) (0)" region="-7" universe="8" />
<cell id="24" material="6" name="GT empty at dashpot grid (intermediate) (1)" region="7 -8" universe="8" />
<cell id="25" material="8" name="GT empty at dashpot grid (intermediate) (last)" region="8 20 -21 22 -23" universe="8" />
<cell id="26" material="6" name="GT empty at dashpot grid (intermediate) (grid)" region="~(20 -21 22 -23)" universe="8" />
<cell id="27" material="8" name="GT empty nozzle (0)" region="-7" universe="9" />
<cell id="28" material="6" name="GT empty nozzle (1)" region="7 -8" universe="9" />
<cell id="29" material="8" name="GT empty nozzle (last)" region="8" universe="9" />
<cell fill="1" id="30" name="GT empty (0)" region="-32" universe="10" />
<cell fill="1" id="31" name="GT empty (1)" region="32 -33" universe="10" />
<cell fill="1" id="32" name="GT empty (2)" region="33 -34" universe="10" />
<cell fill="6" id="33" name="GT empty (3)" region="34 -35" universe="10" />
<cell fill="6" id="34" name="GT empty (4)" region="35 -38" universe="10" />
<cell fill="7" id="35" name="GT empty (5)" region="38 -39" universe="10" />
<cell fill="6" id="36" name="GT empty (6)" region="39 -48" universe="10" />
<cell fill="2" id="37" name="GT empty (7)" region="48 -40" universe="10" />
<cell fill="4" id="38" name="GT empty (8)" region="40 -41" universe="10" />
<cell fill="2" id="39" name="GT empty (9)" region="41 -42" universe="10" />
<cell fill="4" id="40" name="GT empty (10)" region="42 -43" universe="10" />
<cell fill="2" id="41" name="GT empty (11)" region="43 -44" universe="10" />
<cell fill="4" id="42" name="GT empty (12)" region="44 -45" universe="10" />
<cell fill="2" id="43" name="GT empty (13)" region="45 -36" universe="10" />
<cell fill="2" id="44" name="GT empty (14)" region="36 -46" universe="10" />
<cell fill="4" id="45" name="GT empty (15)" region="46 -47" universe="10" />
<cell fill="2" id="46" name="GT empty (16)" region="47 -49" universe="10" />
<cell fill="2" id="47" name="GT empty (17)" region="49 -50" universe="10" />
<cell fill="2" id="48" name="GT empty (18)" region="50 -51" universe="10" />
<cell fill="1" id="49" name="GT empty (19)" region="51 -52" universe="10" />
<cell fill="1" id="50" name="GT empty (last)" region="52" universe="10" />
<cell fill="1" id="51" name="GT empty instr (0)" region="-32" universe="11" />
<cell fill="1" id="52" name="GT empty instr (1)" region="32 -33" universe="11" />
<cell fill="1" id="53" name="GT empty instr (2)" region="33 -34" universe="11" />
<cell fill="10" id="54" name="GT empty instr (3)" region="34 -35" universe="11" />
<cell fill="10" id="55" name="GT empty instr (4)" region="35 -38" universe="11" />
<cell fill="3" id="56" name="GT empty instr (5)" region="38 -39" universe="11" />
<cell fill="10" id="57" name="GT empty instr (6)" region="39 -48" universe="11" />
<cell fill="10" id="58" name="GT empty instr (7)" region="48 -40" universe="11" />
<cell fill="4" id="59" name="GT empty instr (8)" region="40 -41" universe="11" />
<cell fill="10" id="60" name="GT empty instr (9)" region="41 -42" universe="11" />
<cell fill="4" id="61" name="GT empty instr (10)" region="42 -43" universe="11" />
<cell fill="10" id="62" name="GT empty instr (11)" region="43 -44" universe="11" />
<cell fill="4" id="63" name="GT empty instr (12)" region="44 -45" universe="11" />
<cell fill="10" id="64" name="GT empty instr (13)" region="45 -36" universe="11" />
<cell fill="10" id="65" name="GT empty instr (14)" region="36 -46" universe="11" />
<cell fill="4" id="66" name="GT empty instr (15)" region="46 -47" universe="11" />
<cell fill="10" id="67" name="GT empty instr (16)" region="47 -49" universe="11" />
<cell fill="10" id="68" name="GT empty instr (17)" region="49 -50" universe="11" />
<cell fill="10" id="69" name="GT empty instr (18)" region="50 -51" universe="11" />
<cell fill="1" id="70" name="GT empty instr (19)" region="51 -52" universe="11" />
<cell fill="1" id="71" name="GT empty instr (last)" region="52" universe="11" />
<cell id="72" material="2" name="IT (0)" region="-16" universe="12" />
<cell id="73" material="6" name="IT (1)" region="16 -17" universe="12" />
<cell id="74" material="8" name="IT (2)" region="17 -5" universe="12" />
<cell id="75" material="6" name="IT (3)" region="5 -6" universe="12" />
<cell id="76" material="8" name="IT (last)" region="6" universe="12" />
<cell id="77" material="2" name="IT grid (bottom) (0)" region="-16" universe="13" />
<cell id="78" material="6" name="IT grid (bottom) (1)" region="16 -17" universe="13" />
<cell id="79" material="8" name="IT grid (bottom) (2)" region="17 -5" universe="13" />
<cell id="80" material="6" name="IT grid (bottom) (3)" region="5 -6" universe="13" />
<cell id="81" material="8" name="IT grid (bottom) (last)" region="6 20 -21 22 -23" universe="13" />
<cell id="82" material="3" name="IT grid (bottom) (grid)" region="~(20 -21 22 -23)" universe="13" />
<cell id="83" material="2" name="IT grid (intermediate) (0)" region="-16" universe="14" />
<cell id="84" material="6" name="IT grid (intermediate) (1)" region="16 -17" universe="14" />
<cell id="85" material="8" name="IT grid (intermediate) (2)" region="17 -5" universe="14" />
<cell id="86" material="6" name="IT grid (intermediate) (3)" region="5 -6" universe="14" />
<cell id="87" material="8" name="IT grid (intermediate) (last)" region="6 20 -21 22 -23" universe="14" />
<cell id="88" material="6" name="IT grid (intermediate) (grid)" region="~(20 -21 22 -23)" universe="14" />
<cell id="94" material="2" name="IT dashpot (0)" region="-16" universe="16" />
<cell id="95" material="6" name="IT dashpot (1)" region="16 -17" universe="16" />
<cell id="96" material="8" name="IT dashpot (last)" region="17" universe="16" />
<cell fill="16" id="97" name="GT instr (0)" region="-32" universe="17" />
<cell fill="16" id="98" name="GT instr (1)" region="32 -33" universe="17" />
<cell fill="16" id="99" name="GT instr (2)" region="33 -34" universe="17" />
<cell fill="12" id="100" name="GT instr (3)" region="34 -35" universe="17" />
<cell fill="12" id="101" name="GT instr (4)" region="35 -38" universe="17" />
<cell fill="13" id="102" name="GT instr (5)" region="38 -39" universe="17" />
<cell fill="12" id="103" name="GT instr (6)" region="39 -48" universe="17" />
<cell fill="12" id="104" name="GT instr (7)" region="48 -40" universe="17" />
<cell fill="14" id="105" name="GT instr (8)" region="40 -41" universe="17" />
<cell fill="12" id="106" name="GT instr (9)" region="41 -42" universe="17" />
<cell fill="14" id="107" name="GT instr (10)" region="42 -43" universe="17" />
<cell fill="12" id="108" name="GT instr (11)" region="43 -44" universe="17" />
<cell fill="14" id="109" name="GT instr (12)" region="44 -45" universe="17" />
<cell fill="12" id="110" name="GT instr (13)" region="45 -36" universe="17" />
<cell fill="12" id="111" name="GT instr (14)" region="36 -46" universe="17" />
<cell fill="14" id="112" name="GT instr (15)" region="46 -47" universe="17" />
<cell fill="12" id="113" name="GT instr (16)" region="47 -49" universe="17" />
<cell fill="12" id="114" name="GT instr (17)" region="49 -50" universe="17" />
<cell fill="12" id="115" name="GT instr (18)" region="50 -51" universe="17" />
<cell fill="16" id="116" name="GT instr (19)" region="51 -52" universe="17" />
<cell fill="1" id="117" name="GT instr (last)" region="52" universe="17" />
<cell id="118" material="7" name="CR (0)" region="-9" universe="18" />
<cell id="119" material="2" name="CR (1)" region="9 -10" universe="18" />
<cell id="120" material="4" name="CR (2)" region="10 -5" universe="18" />
<cell id="121" material="8" name="CR (3)" region="5 -6" universe="18" />
<cell id="122" material="8" name="CR (last)" region="6" universe="18" />
<cell id="123" material="7" name="CR grid (bottom) (0)" region="-9" universe="19" />
<cell id="124" material="2" name="CR grid (bottom) (1)" region="9 -10" universe="19" />
<cell id="125" material="4" name="CR grid (bottom) (2)" region="10 -5" universe="19" />
<cell id="126" material="8" name="CR grid (bottom) (3)" region="5 -6" universe="19" />
<cell id="127" material="8" name="CR grid (bottom) (last)" region="6 20 -21 22 -23" universe="19" />
<cell id="128" material="3" name="CR grid (bottom) (grid)" region="~(20 -21 22 -23)" universe="19" />
<cell id="129" material="7" name="CR grid (intermediate) (0)" region="-9" universe="20" />
<cell id="130" material="2" name="CR grid (intermediate) (1)" region="9 -10" universe="20" />
<cell id="131" material="4" name="CR grid (intermediate) (2)" region="10 -5" universe="20" />
<cell id="132" material="8" name="CR grid (intermediate) (3)" region="5 -6" universe="20" />
<cell id="133" material="8" name="CR grid (intermediate) (last)" region="6 20 -21 22 -23" universe="20" />
<cell id="134" material="6" name="CR grid (intermediate) (grid)" region="~(20 -21 22 -23)" universe="20" />
<cell id="135" material="7" name="CR nozzle (0)" region="-9" universe="21" />
<cell id="136" material="2" name="CR nozzle (1)" region="9 -10" universe="21" />
<cell id="137" material="4" name="CR nozzle (2)" region="10 -11" universe="21" />
<cell id="138" material="8" name="CR nozzle (last)" region="11" universe="21" />
<cell id="139" material="4" name="CR blank (0)" region="-9" universe="22" />
<cell id="140" material="2" name="CR blank (1)" region="9 -10" universe="22" />
<cell id="141" material="4" name="CR blank (2)" region="10 -11" universe="22" />
<cell id="142" material="8" name="CR blank (3)" region="11 -5" universe="22" />
<cell id="143" material="6" name="CR blank (4)" region="5 -6" universe="22" />
<cell id="144" material="8" name="CR blank (last)" region="6" universe="22" />
<cell id="145" material="4" name="CR blank grid (bottom) (0)" region="-9" universe="23" />
<cell id="146" material="2" name="CR blank grid (bottom) (1)" region="9 -10" universe="23" />
<cell id="147" material="4" name="CR blank grid (bottom) (2)" region="10 -11" universe="23" />
<cell id="148" material="8" name="CR blank grid (bottom) (3)" region="11 -5" universe="23" />
<cell id="149" material="6" name="CR blank grid (bottom) (4)" region="5 -6" universe="23" />
<cell id="150" material="8" name="CR blank grid (bottom) (last)" region="6 20 -21 22 -23" universe="23" />
<cell id="151" material="3" name="CR blank grid (bottom) (grid)" region="~(20 -21 22 -23)" universe="23" />
<cell id="152" material="4" name="CR blank grid (intermediate) (0)" region="-9" universe="24" />
<cell id="153" material="2" name="CR blank grid (intermediate) (1)" region="9 -10" universe="24" />
<cell id="154" material="4" name="CR blank grid (intermediate) (2)" region="10 -11" universe="24" />
<cell id="155" material="8" name="CR blank grid (intermediate) (3)" region="11 -5" universe="24" />
<cell id="156" material="6" name="CR blank grid (intermediate) (4)" region="5 -6" universe="24" />
<cell id="157" material="8" name="CR blank grid (intermediate) (last)" region="6 20 -21 22 -23" universe="24" />
<cell id="158" material="6" name="CR blank grid (intermediate) (grid)" region="~(20 -21 22 -23)" universe="24" />
<cell id="159" material="4" name="CR blank nozzle (0)" region="-9" universe="25" />
<cell id="160" material="2" name="CR blank nozzle (1)" region="9 -10" universe="25" />
<cell id="161" material="4" name="CR blank nozzle (2)" region="10 -11" universe="25" />
<cell id="162" material="8" name="CR blank nozzle (last)" region="11" universe="25" />
<cell id="163" material="4" name="CR blank bare (0)" region="-9" universe="26" />
<cell id="164" material="2" name="CR blank bare (1)" region="9 -10" universe="26" />
<cell id="165" material="4" name="CR blank bare (2)" region="10 -11" universe="26" />
<cell id="166" material="8" name="CR blank bare (last)" region="11" universe="26" />
<cell id="167" material="7" name="CR bare (0)" region="-9" universe="27" />
<cell id="168" material="2" name="CR bare (1)" region="9 -10" universe="27" />
<cell id="169" material="4" name="CR bare (2)" region="10 -11" universe="27" />
<cell id="170" material="8" name="CR bare (last)" region="11" universe="27" />
<cell fill="1" id="309" name="GT CR bank D dummy (0)" region="-34" universe="46" />
<cell fill="6" id="310" name="GT CR bank D dummy (1)" region="34 -48" universe="46" />
<cell fill="10" id="311" name="GT CR bank D dummy (2)" region="48 -70" universe="46" />
<cell fill="18" id="312" name="GT CR bank D dummy (3)" region="70 -69" universe="46" />
<cell fill="22" id="313" name="GT CR bank D dummy (last)" region="69" universe="46" />
<cell fill="1" id="314" name="GT CR bank D dummy grid (bottom) (0)" region="-34" universe="47" />
<cell fill="7" id="315" name="GT CR bank D dummy grid (bottom) (1)" region="34 -48" universe="47" />
<cell fill="3" id="316" name="GT CR bank D dummy grid (bottom) (2)" region="48 -70" universe="47" />
<cell fill="19" id="317" name="GT CR bank D dummy grid (bottom) (3)" region="70 -69" universe="47" />
<cell fill="23" id="318" name="GT CR bank D dummy grid (bottom) (last)" region="69" universe="47" />
<cell fill="1" id="319" name="GT CR bank D dummy grid (intermediate) (0)" region="-34" universe="48" />
<cell fill="8" id="320" name="GT CR bank D dummy grid (intermediate) (1)" region="34 -48" universe="48" />
<cell fill="4" id="321" name="GT CR bank D dummy grid (intermediate) (2)" region="48 -70" universe="48" />
<cell fill="20" id="322" name="GT CR bank D dummy grid (intermediate) (3)" region="70 -69" universe="48" />
<cell fill="24" id="323" name="GT CR bank D dummy grid (intermediate) (last)" region="69" universe="48" />
<cell fill="1" id="324" name="GT CR bank D dummy nozzle (0)" region="-34" universe="49" />
<cell fill="9" id="325" name="GT CR bank D dummy nozzle (1)" region="34 -48" universe="49" />
<cell fill="5" id="326" name="GT CR bank D dummy nozzle (2)" region="48 -70" universe="49" />
<cell fill="21" id="327" name="GT CR bank D dummy nozzle (3)" region="70 -69" universe="49" />
<cell fill="25" id="328" name="GT CR bank D dummy nozzle (last)" region="69" universe="49" />
<cell fill="1" id="329" name="GT CR bank D dummy bare (0)" region="-34" universe="50" />
<cell fill="9" id="330" name="GT CR bank D dummy bare (1)" region="34 -48" universe="50" />
<cell fill="5" id="331" name="GT CR bank D dummy bare (2)" region="48 -70" universe="50" />
<cell fill="27" id="332" name="GT CR bank D dummy bare (3)" region="70 -69" universe="50" />
<cell fill="26" id="333" name="GT CR bank D dummy bare (last)" region="69" universe="50" />
<cell fill="1" id="334" name="GT CR bank D (0)" region="-32" universe="51" />
<cell fill="49" id="335" name="GT CR bank D (1)" region="32 -33" universe="51" />
<cell fill="49" id="336" name="GT CR bank D (2)" region="33 -34" universe="51" />
<cell fill="46" id="337" name="GT CR bank D (3)" region="34 -35" universe="51" />
<cell fill="46" id="338" name="GT CR bank D (4)" region="35 -38" universe="51" />
<cell fill="47" id="339" name="GT CR bank D (5)" region="38 -39" universe="51" />
<cell fill="46" id="340" name="GT CR bank D (6)" region="39 -48" universe="51" />
<cell fill="46" id="341" name="GT CR bank D (7)" region="48 -40" universe="51" />
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<cell fill="46" id="343" name="GT CR bank D (9)" region="41 -42" universe="51" />
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98 98 98 98 98 10 98 98 10 98 98 10 98 98 98 98 98
98 98 98 98 98 98 98 98 98 98 98 98 98 98 98 98 98
98 98 98 98 98 98 98 98 98 98 98 98 98 98 98 98 98 </universes>
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102 102 102 102 102 102 102 102 102 102 102 102 102 102 102 102 102
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102 102 102 102 102 51 102 102 51 102 102 51 102 102 102 102 102
102 102 102 102 102 102 102 102 102 102 102 102 102 102 102 102 102
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106 106 106 106 106 106 106 106 106 106 106 106 106 106 106 106 106
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106 106 106 106 106 106 106 106 106 106 106 106 106 106 106 106 106
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235 235 216 217 218 219 220 235 235
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<surface coeffs="-9.816879130434781 8.49237956521739 0.506426304347826" id="94" type="z-cylinder" />
<surface coeffs="6.155027391304348 -6.155027391304348 0.506426304347826" id="95" type="z-cylinder" />
<surface coeffs="2.726910869565218 1.402411304347826 0.506426304347826" id="96" type="z-cylinder" />
<surface coeffs="-1.5582347826086949 -2.726910869565218 0.506426304347826" id="97" type="z-cylinder" />
<surface coeffs="6.232939130434779 0 0.506426304347826" id="98" type="z-cylinder" />
<surface coeffs="6.232939130434779 5.2200865217391295 0.506426304347826" id="99" type="z-cylinder" />
<surface coeffs="6.232939130434779 10.440173043478259 0.506426304347826" id="100" type="z-cylinder" />
<surface coeffs="6.232939130434779 -5.2200865217391295 0.506426304347826" id="101" type="z-cylinder" />
<surface coeffs="6.232939130434779 -10.440173043478259 0.506426304347826" id="102" type="z-cylinder" />
<surface coeffs="1.5582347826086949 2.6100432608695647 0.506426304347826" id="103" type="z-cylinder" />
<surface coeffs="1.5582347826086949 7.830129782608694 0.506426304347826" id="104" type="z-cylinder" />
<surface coeffs="1.5582347826086949 -2.6100432608695647 0.506426304347826" id="105" type="z-cylinder" />
<surface coeffs="1.5582347826086949 -7.830129782608694 0.506426304347826" id="106" type="z-cylinder" />
<surface coeffs="-2.726910869565218 5.921292173913042 0.506426304347826" id="107" type="z-cylinder" />
<surface coeffs="-2.726910869565218 -5.921292173913042 0.506426304347826" id="108" type="z-cylinder" />
<surface coeffs="6.232939130434779 -11.063466956521738 0.506426304347826" id="109" type="z-cylinder" />
<surface coeffs="6.232939130434779 -5.84338043478261 0.506426304347826" id="110" type="z-cylinder" />
<surface coeffs="6.232939130434779 -0.6232939130434794 0.506426304347826" id="111" type="z-cylinder" />
<surface coeffs="6.232939130434779 4.596792608695651 0.506426304347826" id="112" type="z-cylinder" />
<surface coeffs="6.232939130434779 9.816879130434778 0.506426304347826" id="113" type="z-cylinder" />
<surface coeffs="1.5582347826086949 -8.453423695652173 0.506426304347826" id="114" type="z-cylinder" />
<surface coeffs="1.5582347826086949 -3.2333371739130428 0.506426304347826" id="115" type="z-cylinder" />
<surface coeffs="1.5582347826086949 7.206835869565214 0.506426304347826" id="116" type="z-cylinder" />
<surface coeffs="-0.15582347826087073 1.6361465217391302 0.506426304347826" id="117" type="z-cylinder" />
<surface coeffs="-3.5060282608695648 -7.089968260869565 0.506426304347826" id="118" type="z-cylinder" />
<surface coeffs="6.232939130434779 11.063466956521738 0.506426304347826" id="119" type="z-cylinder" />
<surface coeffs="6.232939130434779 5.84338043478261 0.506426304347826" id="120" type="z-cylinder" />
<surface coeffs="6.232939130434779 0.6232939130434794 0.506426304347826" id="121" type="z-cylinder" />
<surface coeffs="6.232939130434779 -4.596792608695651 0.506426304347826" id="122" type="z-cylinder" />
<surface coeffs="6.232939130434779 -9.816879130434778 0.506426304347826" id="123" type="z-cylinder" />
<surface coeffs="1.5582347826086949 8.453423695652173 0.506426304347826" id="124" type="z-cylinder" />
<surface coeffs="1.5582347826086949 3.2333371739130428 0.506426304347826" id="125" type="z-cylinder" />
<surface coeffs="1.5582347826086949 -7.206835869565214 0.506426304347826" id="126" type="z-cylinder" />
<surface coeffs="-0.15582347826087073 -1.6361465217391302 0.506426304347826" id="127" type="z-cylinder" />
<surface coeffs="-3.5060282608695648 7.089968260869565 0.506426304347826" id="128" type="z-cylinder" />
<surface coeffs="4.674704347826085 -3.1164695652173915 0.506426304347826" id="129" type="z-cylinder" />
</geometry>

75
smr/fresh/tallies.xml
View file

@ -1,75 +0,0 @@
<?xml version='1.0' encoding='utf-8'?>
<tallies>
<filter id="1" type="distribcell">
<bins>684</bins>
</filter>
<filter id="2" type="distribcell">
<bins>688</bins>
</filter>
<filter id="3" type="distribcell">
<bins>693</bins>
</filter>
<filter id="4" type="distribcell">
<bins>719</bins>
</filter>
<filter id="5" type="distribcell">
<bins>723</bins>
</filter>
<filter id="6" type="distribcell">
<bins>728</bins>
</filter>
<filter id="7" type="distribcell">
<bins>754</bins>
</filter>
<filter id="8" type="distribcell">
<bins>758</bins>
</filter>
<filter id="9" type="distribcell">
<bins>763</bins>
</filter>
<tally id="1" name="depletion tally">
<filters>1</filters>
<nuclides>O16 O17 U234 U235 U238 U236</nuclides>
<scores>(n,p) (n,a) (n,gamma) fission (n,2n) (n,3n) (n,4n)</scores>
</tally>
<tally id="2" name="depletion tally">
<filters>2</filters>
<nuclides>O16 O17 U234 U235 U238 U236</nuclides>
<scores>(n,p) (n,a) (n,gamma) fission (n,2n) (n,3n) (n,4n)</scores>
</tally>
<tally id="3" name="depletion tally">
<filters>3</filters>
<nuclides>O16 O17 U234 U235 U238 U236</nuclides>
<scores>(n,p) (n,a) (n,gamma) fission (n,2n) (n,3n) (n,4n)</scores>
</tally>
<tally id="4" name="depletion tally">
<filters>4</filters>
<nuclides>O16 O17 U234 U235 U238 U236</nuclides>
<scores>(n,p) (n,a) (n,gamma) fission (n,2n) (n,3n) (n,4n)</scores>
</tally>
<tally id="5" name="depletion tally">
<filters>5</filters>
<nuclides>O16 O17 U234 U235 U238 U236</nuclides>
<scores>(n,p) (n,a) (n,gamma) fission (n,2n) (n,3n) (n,4n)</scores>
</tally>
<tally id="6" name="depletion tally">
<filters>6</filters>
<nuclides>O16 O17 U234 U235 U238 U236</nuclides>
<scores>(n,p) (n,a) (n,gamma) fission (n,2n) (n,3n) (n,4n)</scores>
</tally>
<tally id="7" name="depletion tally">
<filters>7</filters>
<nuclides>O16 O17 U234 U235 U238 U236</nuclides>
<scores>(n,p) (n,a) (n,gamma) fission (n,2n) (n,3n) (n,4n)</scores>
</tally>
<tally id="8" name="depletion tally">
<filters>8</filters>
<nuclides>O16 O17 U234 U235 U238 U236</nuclides>
<scores>(n,p) (n,a) (n,gamma) fission (n,2n) (n,3n) (n,4n)</scores>
</tally>
<tally id="9" name="depletion tally">
<filters>9</filters>
<nuclides>O16 O17 U234 U235 U238 U236</nuclides>
<scores>(n,p) (n,a) (n,gamma) fission (n,2n) (n,3n) (n,4n)</scores>
</tally>
</tallies>

20
smr/milestones.md Normal file
View file

@ -0,0 +1,20 @@
# Milestone Models
- **AD-SE-08-61, Coupled Multiphysics Driver Implementation** --- This milestone
used the singlerod short/long problems. Generating the model was done with the
script `tests/singlerod/make_openmc_model.py` from the ENRICO repository
(there is a `--short` command line option to generate the short version)
- **AD-SE-08-66, Coupled Assembly Analysis** --- This milestone used the
assembly short and long (v2) problems. Generating the model was done with
`smr/build-assembly-long.py -a 100 --clone` on the ecp-benchmarks repository
(git commit `631fefe`, after pull request #14). In these models, materials are
fully differentiated across each fuel ring/axial segment.
- **AD-SE-08-73, Full core coupled-physics simulation** --- This milestone used
the core-short and core-long (90 layer) models. Generating the models was done
with `smr/build-core-short.py` and `smr/build-assembly-long.py -a 90` on the
ecp-benchmarks repository (git commit `c7b89db`, after pull request #16). In
these models, materials are not differentiated and no grid spacers are
present. The lattice pitch is modified to be exactly 17 times the pin pitch
(slightly different than NuScale specification).

6
smr/smr/__init__.py
View file

@ -0,0 +1,6 @@
_MPA_PER_PSIA = 0.0068947572931683625
# Values from ML17013A274, Table 4.1-1
inlet_temperature = (497 - 32)*5/9 + 273.15
core_average_temperature = (543 - 32)*5/9 + 273.15
system_pressure = 1850*_MPA_PER_PSIA

413
smr/smr/assemblies.py
View file

@ -6,7 +6,7 @@ import openmc
from .materials import mats
from .surfaces import surfs, pin_pitch
from .pins import univs
from .pins import pin_universes
def make_assembly(name, universes):
@ -30,12 +30,12 @@ def make_assembly(name, universes):
# Instantiate the lattice
lattice = openmc.RectLattice(name=name)
lattice.lower_left = [-17.*pin_pitch/2., -17.*pin_pitch/2.]
lattice.pitch = [pin_pitch, pin_pitch]
lattice.lower_left = (-17.*pin_pitch/2., -17.*pin_pitch/2.)
lattice.pitch = (pin_pitch, pin_pitch)
lattice.universes = universes
# Create rectangular prism for lattice grid box
lat_grid_box = (surfs['lat grid box outer'] & ~surfs['lat grid box inner'])
lat_grid_box = surfs['lat grid box outer'] & ~surfs['lat grid box inner']
# Add lattice to bounding cell
univ_name = name + ' lattice'
@ -47,7 +47,7 @@ def make_assembly(name, universes):
# Add outer water cell
cell = openmc.Cell(name=univ_name + ' outer water')
cell.fill = univs['water pin']
cell.fill = mats['H2O']
cell.region = ~surfs['lat grid box outer']
universe.add_cell(cell)
@ -59,7 +59,7 @@ def make_assembly(name, universes):
# Make axial cell for outside of assembly (with sleeve)
cell = openmc.Cell(name=univ_name + ' axial (1)')
cell.fill = mats['SS']
cell.fill = mats['In']
cell.region = lat_grid_box & +surfs['grid1bot'] & -surfs['grid1top']
universe.add_cell(cell)
@ -95,290 +95,171 @@ def make_assembly(name, universes):
# Make axial cell for outside of assembly (with sleeve)
cell = openmc.Cell(name=univ_name + ' axial (7)')
cell.fill = mats['SS']
cell.fill = mats['Zr']
cell.region = lat_grid_box & +surfs['grid4bot'] & -surfs['grid4top']
universe.add_cell(cell)
# Make top axial cell for outside of assembly (without sleeve)
cell = openmc.Cell(name=univ_name + ' axial (8)')
cell.fill = mats['H2O']
cell.region = lat_grid_box & +surfs['grid4top'] & -surfs['grid5bot']
universe.add_cell(cell)
# Make axial cell for outside of assembly (with sleeve)
cell = openmc.Cell(name=univ_name + ' axial (9)')
cell.fill = mats['Zr']
cell.region = lat_grid_box & +surfs['grid5bot'] & -surfs['grid5top']
universe.add_cell(cell)
# Make top axial cell for outside of assembly (without sleeve)
cell = openmc.Cell(name=univ_name + ' axial (last)')
cell.fill = mats['H2O']
cell.region = lat_grid_box & +surfs['grid4top']
cell.region = lat_grid_box & +surfs['grid5top']
universe.add_cell(cell)
return universe
# commonly needed universes
gtu = univs['GT empty']
gti = univs['GT empty instr']
bas = univs['BA stack']
ins = univs['IT stack']
crA = univs['GT CR bank A']
crB = univs['GT CR bank B']
crC = univs['GT CR bank C']
crD = univs['GT CR bank D']
crSA = univs['GT CR bank SA']
crSB = univs['GT CR bank SB']
crSC = univs['GT CR bank SC']
crSD = univs['GT CR bank SD']
crSE = univs['GT CR bank SE']
def assembly_universes(ring_radii, num_axial, depleted):
"""Generate universes for SMR fuel assemblies.
Parameters
----------
ring_radii : iterable of float
Radii of rings in fuel (note that this doesn't need to include the
full fuel pin radius)
num_axial : int
Number of axial subdivisions in fuel
depleted : bool
Whether fuel should contain nuclides as though it were depleted
Returns
-------
dict
Dictionary mapping a universe name to a openmc.Universe object
"""
pins = pin_universes(ring_radii, num_axial, depleted)
# Create dictionary to store assembly universes
univs = {}
# commonly needed universes
gtu = pins['GT empty stack']
gti = pins['GT empty instr']
ins = pins['IT stack']
crA = pins['GT CR bank A']
crB = pins['GT CR bank B']
crC = pins['GT CR bank C']
crD = pins['GT CR bank D']
crSA = pins['GT CR bank SA']
crSB = pins['GT CR bank SB']
crSC = pins['GT CR bank SC']
crSD = pins['GT CR bank SD']
crSE = pins['GT CR bank SE']
# Define the NumPy array indices for assembly locations where there
# may be CR guide tubes, instrument tubes and burnable absorbers
nonfuel_y = \
np.array([2,2,2,3,3,5,5,5,5,5,8,8,8,8,8,11,11,11,11,11,13,13,14,14,14])
nonfuel_x = \
np.array([5,8,11,3,13,2,5,8,11,14,2,5,8,11,14,2,5,8,11,14,3,13,5,8,11])
# Define the NumPy array indices for assembly locations where there
# may be CR guide tubes, instrument tubes and burnable absorbers
nonfuel_y = \
np.array([2,2,2,3,3,5,5,5,5,5,8,8,8,8,8,11,11,11,11,11,13,13,14,14,14])
nonfuel_x = \
np.array([5,8,11,3,13,2,5,8,11,14,2,5,8,11,14,2,5,8,11,14,3,13,5,8,11])
#### 1.6% ENRICHED ASSEMBLIES
#### 1.6% ENRICHED ASSEMBLIES
for cent, comment in [(gti, ''), (ins, ' instr')]:
# NO BURNABLE ABSORBERS
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (1.6%) stack']
universes[nonfuel_y, nonfuel_x] = [ gtu, gtu, gtu,
gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu, cent, gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu,
gtu, gtu, gtu ]
univs['Assembly (1.6%)' + comment] = \
make_assembly('Assembly (1.6%) no BAs' + comment, universes)
# WITH EACH CONTROL ROD BANK
for bank, comment2 in [(crA, 'A'), (crB, 'B'), (crC, 'C'), (crD, 'D'),
(crSB, 'SB'), (crSC, 'SC'), (crSD, 'SD'), (crSE, 'SE')]:
for cent, comment in [(gti, ''), (ins, ' instr')]:
# NO BURNABLE ABSORBERS
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (1.6%) stack']
universes[nonfuel_y, nonfuel_x] = [ bank, bank, bank,
bank, bank,
bank, bank, bank, bank, bank,
bank, bank, cent, bank, bank,
bank, bank, bank, bank, bank,
bank, bank,
bank, bank, bank ]
univs['Assembly (1.6%) CR {}'.format(comment2) + comment] = \
make_assembly('Assembly (1.6%) CR {}'.format(comment2) + comment, universes)
universes[:,:] = pins['Fuel (1.6%) stack']
universes[nonfuel_y, nonfuel_x] = [ gtu, gtu, gtu,
gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu, cent, gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu,
gtu, gtu, gtu ]
univs['Assembly (1.6%)' + comment] = \
make_assembly('Assembly (1.6%) no BAs' + comment, universes)
# WITH EACH CONTROL ROD BANK
for bank, comment2 in [(crA, 'A'), (crB, 'B'), (crC, 'C'), (crD, 'D'),
(crSB, 'SB'), (crSC, 'SC'), (crSD, 'SD'), (crSE, 'SE')]:
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = pins['Fuel (1.6%) stack']
universes[nonfuel_y, nonfuel_x] = [ bank, bank, bank,
bank, bank,
bank, bank, bank, bank, bank,
bank, bank, cent, bank, bank,
bank, bank, bank, bank, bank,
bank, bank,
bank, bank, bank ]
univs['Assembly (1.6%) CR {}'.format(comment2) + comment] = \
make_assembly('Assembly (1.6%) CR {}'.format(comment2) + comment, universes)
#### 2.4% ENRICHED ASSEMBLIES
#### 2.4% ENRICHED ASSEMBLIES
for cent, comment in [(gti, ''), (ins, ' instr')]:
for cent, comment in [(gti, ''), (ins, ' instr')]:
# NO BURNABLE ABSORBERS
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (2.4%) stack']
universes[nonfuel_y, nonfuel_x] = [ gtu, gtu, gtu,
gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu, cent, gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu,
gtu, gtu, gtu ]
univs['Assembly (2.4%) no BAs' + comment] = \
make_assembly('Assembly (2.4%) no BAs' + comment, universes)
# NO BURNABLE ABSORBERS
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = pins['Fuel (2.4%) stack']
universes[nonfuel_y, nonfuel_x] = [ gtu, gtu, gtu,
gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu, cent, gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu,
gtu, gtu, gtu ]
univs['Assembly (2.4%)' + comment] = \
make_assembly('Assembly (2.4%) no BAs' + comment, universes)
# WITH CONTROL ROD D BANK
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (2.4%) stack']
universes[nonfuel_y, nonfuel_x] = [ crD, crD, crD,
crD, crD,
crD, crD, crD, crD, crD,
crD, crD, cent, crD, crD,
crD, crD, crD, crD, crD,
crD, crD,
crD, crD, crD ]
univs['Assembly (2.4%) CR D' + comment] = \
make_assembly('Assembly (2.4%) CR D' + comment, universes)
# WITH CONTROL ROD D BANK
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = pins['Fuel (2.4%) stack']
universes[nonfuel_y, nonfuel_x] = [ crD, crD, crD,
crD, crD,
crD, crD, crD, crD, crD,
crD, crD, cent, crD, crD,
crD, crD, crD, crD, crD,
crD, crD,
crD, crD, crD ]
univs['Assembly (2.4%) CR D' + comment] = \
make_assembly('Assembly (2.4%) CR D' + comment, universes)
# WITH 12 BURNABLE ABSORBERS
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (2.4%) stack']
universes[nonfuel_y, nonfuel_x] = [ bas, gtu, bas,
bas, bas,
bas, gtu, gtu, gtu, bas,
gtu, gtu, cent, gtu, gtu,
bas, gtu, gtu, gtu, bas,
bas, bas,
bas, gtu, bas ]
univs['Assembly (2.4%) 12BA' + comment] = \
make_assembly('Assembly (2.4%) 12BA' + comment, universes)
#### 3.1% ENRICHED ASSEMBLIES
# WITH 16 BURNABLE ABSORBERS
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (2.4%) stack']
universes[nonfuel_y, nonfuel_x] = [ bas, bas, bas,
bas, bas,
bas, gtu, gtu, gtu, bas,
bas, gtu, cent, gtu, bas,
bas, gtu, gtu, gtu, bas,
bas, bas,
bas, bas, bas ]
univs['Assembly (2.4%) 16BA' + comment] = \
make_assembly('Assembly (2.4%) 16BA' + comment, universes)
for cent, comment in [(gti, ''), (ins, ' instr')]:
# NO BURNABLE ABSORBERS
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = pins['Fuel (3.1%) stack']
universes[nonfuel_y, nonfuel_x] = [ gtu, gtu, gtu,
gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu, cent, gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu,
gtu, gtu, gtu ]
univs['Assembly (3.1%)' + comment] = \
make_assembly('Assembly (3.1%) no BAs' + comment, universes)
# WITH CONTROL ROD SA BANK
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = pins['Fuel (3.1%) stack']
universes[nonfuel_y, nonfuel_x] = [ crSA, crSA, crSA,
crSA, crSA,
crSA, crSA, crSA, crSA, crSA,
crSA, crSA, cent, crSA, crSA,
crSA, crSA, crSA, crSA, crSA,
crSA, crSA,
crSA, crSA, crSA ]
univs['Assembly (3.1%) CR SA' + comment] = \
make_assembly('Assembly (3.1%) CR SA' + comment, universes)
#### 3.1% ENRICHED ASSEMBLIES
for cent, comment in [(gti, ''), (ins, ' instr')]:
# NO BURNABLE ABSORBERS
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (3.1%) stack']
universes[nonfuel_y, nonfuel_x] = [ gtu, gtu, gtu,
gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu, cent, gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu,
gtu, gtu, gtu ]
univs['Assembly (3.1%)' + comment] = \
make_assembly('Assembly (3.1%) no BAs' + comment, universes)
# WITH CONTROL ROD SA BANK
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (3.1%) stack']
universes[nonfuel_y, nonfuel_x] = [ crSA, crSA, crSA,
crSA, crSA,
crSA, crSA, crSA, crSA, crSA,
crSA, crSA, cent, crSA, crSA,
crSA, crSA, crSA, crSA, crSA,
crSA, crSA,
crSA, crSA, crSA ]
univs['Assembly (3.1%) CR SA' + comment] = \
make_assembly('Assembly (3.1%) CR SA' + comment, universes)
# WITH 20 BURNABLE ABSORBERS
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (3.1%) stack']
universes[nonfuel_y, nonfuel_x] = [ bas, bas, bas,
bas, bas,
bas, bas, gtu, bas, bas,
bas, gtu, cent, gtu, bas,
bas, bas, gtu, bas, bas,
bas, bas,
bas, bas, bas ]
univs['Assembly (3.1%) 20BA' + comment] = \
make_assembly('Assembly (3.1%) 20BA' + comment, universes)
# WITH 16 BURNABLE ABSORBERS
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (3.1%) stack']
universes[nonfuel_y, nonfuel_x] = [ bas, bas, bas,
bas, bas,
bas, gtu, gtu, gtu, bas,
bas, gtu, cent, gtu, bas,
bas, gtu, gtu, gtu, bas,
bas, bas,
bas, bas, bas ]
univs['Assembly (3.1%) 16BA' + comment] = \
make_assembly('Assembly (3.1%) 16BA' + comment, universes)
# WITH 15 BURNABLE ABSORBERS NW
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (3.1%) stack']
universes[nonfuel_y, nonfuel_x] = [ gtu, gtu, gtu,
gtu, gtu,
gtu, bas, bas, bas, bas,
gtu, bas, cent, bas, bas,
gtu, bas, bas, bas, bas,
gtu, bas,
bas, bas, bas ]
univs['Assembly (3.1%) 15BANW' + comment] = \
make_assembly('Assembly (3.1%) 15BANW' + comment, universes)
# WITH 15 BURNABLE ABSORBERS NE
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (3.1%) stack']
universes[nonfuel_y, nonfuel_x] = [ gtu, gtu, gtu,
gtu, gtu,
bas, bas, bas, bas, gtu,
bas, bas, cent, bas, gtu,
bas, bas, bas, bas, gtu,
bas, gtu,
bas, bas, bas ]
univs['Assembly (3.1%) 15BANE' + comment] = \
make_assembly('Assembly (3.1%) 15BANE' + comment, universes)
# WITH 15 BURNABLE ABSORBERS SW
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (3.1%) stack']
universes[nonfuel_y, nonfuel_x] = [ bas, bas, bas,
gtu, bas,
gtu, bas, bas, bas, bas,
gtu, bas, cent, bas, bas,
gtu, bas, bas, bas, bas,
gtu, gtu,
gtu, gtu, gtu ]
univs['Assembly (3.1%) 15BASW' + comment] = \
make_assembly('Assembly (3.1%) 15BASW' + comment, universes)
# WITH 15 BURNABLE ABSORBERS SE
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (3.1%) stack']
universes[nonfuel_y, nonfuel_x] = [ bas, bas, bas,
bas, gtu,
bas, bas, bas, bas, gtu,
bas, bas, cent, bas, gtu,
bas, bas, bas, bas, gtu,
gtu, gtu,
gtu, gtu, gtu ]
univs['Assembly (3.1%) 15BASE' + comment] = \
make_assembly('Assembly (3.1%) 15BASE' + comment, universes)
# WITH 6 BURNABLE ABSORBERS N
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (3.1%) stack']
universes[nonfuel_y, nonfuel_x] = [ gtu, gtu, gtu,
gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu, cent, gtu, gtu,
bas, gtu, gtu, gtu, bas,
bas, bas,
bas, gtu, bas ]
univs['Assembly (3.1%) 6BAN' + comment] = \
make_assembly('Assembly (3.1%) 6BAN' + comment, universes)
# WITH 6 BURNABLE ABSORBERS S
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (3.1%) stack']
universes[nonfuel_y, nonfuel_x] = [ bas, gtu, bas,
bas, bas,
bas, gtu, gtu, gtu, bas,
gtu, gtu, cent, gtu, gtu,
gtu, gtu, gtu, gtu, gtu,
gtu, gtu,
gtu, gtu, gtu ]
univs['Assembly (3.1%) 6BAS' + comment] = \
make_assembly('Assembly (3.1%) 6BAS' + comment, universes)
# WITH 6 BURNABLE ABSORBERS W
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (3.1%) stack']
universes[nonfuel_y, nonfuel_x] = [ gtu, gtu, bas,
gtu, bas,
gtu, gtu, gtu, gtu, bas,
gtu, gtu, cent, gtu, gtu,
gtu, gtu, gtu, gtu, bas,
gtu, bas,
gtu, gtu, bas ]
univs['Assembly (3.1%) 6BAW' + comment] = \
make_assembly('Assembly (3.1%) 6BAW' + comment, universes)
# WITH 6 BURNABLE ABSORBERS E
universes = np.empty((17,17), dtype=openmc.Universe)
universes[:,:] = univs['Fuel (3.1%) stack']
universes[nonfuel_y, nonfuel_x] = [ bas, gtu, gtu,
bas, gtu,
bas, gtu, gtu, gtu, gtu,
gtu, gtu, cent, gtu, gtu,
bas, gtu, gtu, gtu, gtu,
bas, gtu,
bas, gtu, gtu ]
univs['Assembly (3.1%) 6BAE' + comment] = \
make_assembly('Assembly (3.1%) 6BAE' + comment, universes)
return univs

303
smr/smr/core.py
View file

@ -5,203 +5,150 @@ import numpy as np
import openmc
from .materials import mats
from .surfaces import surfs, lattice_pitch
from .reflector import univs
from .reflector import reflector_universes
from .assemblies import assembly_universes
from smr import surfaces
#### CONSTRUCT MAIN CORE LATTICE
def core_geometry(ring_radii, num_axial, depleted):
"""Generate full core SMR geometry.
core = openmc.RectLattice(name='Main core')
core.lower_left = [-9*lattice_pitch/2, -9*lattice_pitch/2]
core.pitch = [lattice_pitch, lattice_pitch]
universes = np.tile(univs['heavy reflector'], (9,9))
Parameters
----------
ring_radii : iterable of float
Radii of rings in fuel (note that this doesn't need to include the
full fuel pin radius)
num_axial : int
Number of axial subdivisions in fuel
depleted : bool
Whether fuel should contain nuclides as though it were depleted
universes[0, 2] = univs['heavy reflector 0,2']
universes[0, 3] = univs['heavy reflector 0,3']
universes[0, 4] = univs['heavy reflector 0,4']
universes[0, 5] = univs['heavy reflector 0,5']
universes[0, 6] = univs['heavy reflector 0,6']
Returns
-------
openmc.Geometry
SMR full core geometry
universes[1, 1] = univs['heavy reflector 1,1']
universes[1, 2] = univs['heavy reflector NW']
universes[1, 3] = univs['Assembly (3.1%) instr']
universes[1, 4] = univs['Assembly (2.4%) CR D']
universes[1, 5] = univs['Assembly (3.1%) instr']
universes[1, 6] = univs['heavy reflector NE']
universes[1, 7] = univs['heavy reflector 1,7']
"""
assembly = assembly_universes(ring_radii, num_axial, depleted)
reflector = reflector_universes()
universes[2, 0] = univs['heavy reflector 2,0']
universes[2, 1] = univs['heavy reflector NW']
universes[2, 2] = univs['Assembly (3.1%) instr']
universes[2, 3] = univs['Assembly (2.4%) CR D']
universes[2, 4] = univs['Assembly (3.1%) 16BA']
universes[2, 5] = univs['Assembly (2.4%) CR D']
universes[2, 6] = univs['Assembly (3.1%) instr']
universes[2, 7] = univs['heavy reflector NE']
universes[2, 8] = univs['heavy reflector 2,8']
# Construct main core lattice
core = openmc.RectLattice(name='Main core')
lattice_pitch = surfaces.lattice_pitch
core.lower_left = (-9*lattice_pitch/2, -9*lattice_pitch/2)
core.pitch = (lattice_pitch, lattice_pitch)
universes = np.tile(reflector['solid'], (9, 9))
universes[3, 0] = univs['heavy reflector 3,0']
universes[3, 1] = univs['Assembly (3.1%) instr']
universes[3, 2] = univs['Assembly (2.4%) CR D']
universes[3, 3] = univs['Assembly (3.1%) 16BA']
universes[3, 4] = univs['Assembly (2.4%) CR D']
universes[3, 5] = univs['Assembly (3.1%) 16BA']
universes[3, 6] = univs['Assembly (2.4%) CR D']
universes[3, 7] = univs['Assembly (3.1%) instr']
universes[3, 8] = univs['heavy reflector 3,8']
universes[0, 2] = reflector['0,2']
universes[0, 3] = reflector['0,3']
universes[0, 4] = reflector['0,4']
universes[0, 5] = reflector['0,5']
universes[0, 6] = reflector['0,6']
universes[4, 0] = univs['heavy reflector 4,0']
universes[4, 1] = univs['Assembly (2.4%) CR D']
universes[4, 2] = univs['Assembly (3.1%) 16BA']
universes[4, 3] = univs['Assembly (2.4%) CR D']
universes[4, 4] = univs['Assembly (1.6%) instr']
universes[4, 5] = univs['Assembly (2.4%) CR D']
universes[4, 6] = univs['Assembly (3.1%) 16BA']
universes[4, 7] = univs['Assembly (2.4%) CR D']
universes[4, 8] = univs['heavy reflector 4,8']
universes[1, 1] = reflector['1,1']
universes[1, 2] = reflector['NW']
universes[1, 3] = assembly['Assembly (3.1%)']
universes[1, 4] = assembly['Assembly (3.1%)']
universes[1, 5] = assembly['Assembly (3.1%)']
universes[1, 6] = reflector['NE']
universes[1, 7] = reflector['1,7']
universes[5, 0] = univs['heavy reflector 5,0']
universes[5, 1] = univs['Assembly (3.1%) instr']
universes[5, 2] = univs['Assembly (2.4%) CR D']
universes[5, 3] = univs['Assembly (3.1%) 16BA']
universes[5, 4] = univs['Assembly (2.4%) CR D']
universes[5, 5] = univs['Assembly (3.1%) 16BA']
universes[5, 6] = univs['Assembly (2.4%) CR D']
universes[5, 7] = univs['Assembly (3.1%) instr']
universes[5, 8] = univs['heavy reflector 5,8']
universes[2, 0] = reflector['2,0']
universes[2, 1] = reflector['NW']
universes[2, 2] = assembly['Assembly (3.1%)']
universes[2, 3] = assembly['Assembly (2.4%)']
universes[2, 4] = assembly['Assembly (1.6%)']
universes[2, 5] = assembly['Assembly (2.4%)']
universes[2, 6] = assembly['Assembly (3.1%)']
universes[2, 7] = reflector['NE']
universes[2, 8] = reflector['2,8']
universes[6, 0] = univs['heavy reflector 6,0']
universes[6, 1] = univs['heavy reflector SW']
universes[6, 2] = univs['Assembly (3.1%) instr']
universes[6, 3] = univs['Assembly (2.4%) CR D']
universes[6, 4] = univs['Assembly (3.1%) 16BA']
universes[6, 5] = univs['Assembly (2.4%) CR D']
universes[6, 6] = univs['Assembly (3.1%) instr']
universes[6, 7] = univs['heavy reflector SE']
universes[6, 8] = univs['heavy reflector 6,8']
universes[3, 0] = reflector['3,0']
universes[3, 1] = assembly['Assembly (3.1%)']
universes[3, 2] = assembly['Assembly (2.4%)']
universes[3, 3] = assembly['Assembly (1.6%)']
universes[3, 4] = assembly['Assembly (1.6%)']
universes[3, 5] = assembly['Assembly (1.6%)']
universes[3, 6] = assembly['Assembly (2.4%)']
universes[3, 7] = assembly['Assembly (3.1%)']
universes[3, 8] = reflector['3,8']
universes[7, 1] = univs['heavy reflector 7,1']
universes[7, 2] = univs['heavy reflector SW']
universes[7, 3] = univs['Assembly (3.1%) instr']
universes[7, 4] = univs['Assembly (2.4%) CR D']
universes[7, 5] = univs['Assembly (3.1%) instr']
universes[7, 6] = univs['heavy reflector SE']
universes[7, 7] = univs['heavy reflector 7,7']
universes[4, 0] = reflector['4,0']
universes[4, 1] = assembly['Assembly (3.1%)']
universes[4, 2] = assembly['Assembly (1.6%)']
universes[4, 3] = assembly['Assembly (1.6%)']
universes[4, 4] = assembly['Assembly (2.4%)']
universes[4, 5] = assembly['Assembly (1.6%)']
universes[4, 6] = assembly['Assembly (1.6%)']
universes[4, 7] = assembly['Assembly (3.1%)']
universes[4, 8] = reflector['4,8']
universes[8, 2] = univs['heavy reflector 8,2']
universes[8, 3] = univs['heavy reflector 8,3']
universes[8, 4] = univs['heavy reflector 8,4']
universes[8, 5] = univs['heavy reflector 8,5']
universes[8, 6] = univs['heavy reflector 8,6']
universes[5, 0] = reflector['5,0']
universes[5, 1] = assembly['Assembly (3.1%)']
universes[5, 2] = assembly['Assembly (2.4%)']
universes[5, 3] = assembly['Assembly (1.6%)']
universes[5, 4] = assembly['Assembly (1.6%)']
universes[5, 5] = assembly['Assembly (1.6%)']
universes[5, 6] = assembly['Assembly (2.4%)']
universes[5, 7] = assembly['Assembly (3.1%)']
universes[5, 8] = reflector['5,8']
core.universes = universes
universes[6, 0] = reflector['6,0']
universes[6, 1] = reflector['SW']
universes[6, 2] = assembly['Assembly (3.1%)']
universes[6, 3] = assembly['Assembly (2.4%)']
universes[6, 4] = assembly['Assembly (1.6%)']
universes[6, 5] = assembly['Assembly (2.4%)']
universes[6, 6] = assembly['Assembly (3.1%)']
universes[6, 7] = reflector['SE']
universes[6, 8] = reflector['6,8']
universes[7, 1] = reflector['7,1']
universes[7, 2] = reflector['SW']
universes[7, 3] = assembly['Assembly (3.1%)']
universes[7, 4] = assembly['Assembly (3.1%)']
universes[7, 5] = assembly['Assembly (3.1%)']
universes[7, 6] = reflector['SE']
universes[7, 7] = reflector['7,7']
#### CONSTRUCT ROOT UNIVERSE AND CELLS
universes[8, 2] = reflector['8,2']
universes[8, 3] = reflector['8,3']
universes[8, 4] = reflector['8,4']
universes[8, 5] = reflector['8,5']
universes[8, 6] = reflector['8,6']
root_univ = openmc.Universe(universe_id=0, name='root universe')
core.universes = universes
cell = openmc.Cell(name='Main core')
cell.fill = core
cell.region = \
-surfs['core barrel IR'] & +surfs['lower bound'] & -surfs['upper bound']
root_univ.add_cell(cell)
root_univ = openmc.Universe(universe_id=0, name='root universe')
surfs = surfaces.surfs
# Cylinder filled with core lattice
cell = openmc.Cell(name='Main core')
cell.fill = core
cell.region = \
-surfs['core barrel IR'] & +surfs['lower bound'] & -surfs['upper bound']
root_univ.add_cell(cell)
# CONSTRUCT CORE BARREL
# Core barrel
cell = openmc.Cell(name='core barrel')
cell.fill = mats['SS']
cell.region = (+surfs['core barrel IR'] & -surfs['core barrel OR'] &
+surfs['lower bound'] & -surfs['upper bound'])
root_univ.add_cell(cell)
cell = openmc.Cell(name='core barrel')
cell.fill = mats['SS']
cell.region = (+surfs['core barrel IR'] & -surfs['core barrel OR'] &
+surfs['lower bound'] & -surfs['upper bound'])
root_univ.add_cell(cell)
# Downcomer
cell = openmc.Cell(name='downcomer')
cell.fill = mats['H2O']
cell.region = (+surfs['core barrel OR'] & -surfs['RPV IR'] &
+surfs['lower bound'] & -surfs['upper bound'])
root_univ.add_cell(cell)
# Reactor pressure vessel
cell = openmc.Cell(name='reactor pressure vessel')
cell.fill = mats['CS']
cell.region = (+surfs['RPV IR'] & -surfs['RPV OR'] &
+surfs['lower bound'] & -surfs['upper bound'])
root_univ.add_cell(cell)
#### CONSTRUCT NEUTRON SHIELD PANELS
cell = openmc.Cell(name='neutron shield panel NW')
cell.fill = mats['SS']
cell.region = (+surfs['core barrel OR'] & -surfs['neutron shield OR'] &
+surfs['neutron shield NWbot SEtop'] &
-surfs['neutron shield NWtop SEbot'] &
+surfs['lower bound'] & -surfs['upper bound'])
root_univ.add_cell(cell)
cell = openmc.Cell(name='neutron shield panel N')
cell.fill = mats['H2O']
cell.region = (+surfs['core barrel OR'] & -surfs['neutron shield OR'] &
+surfs['neutron shield NWtop SEbot'] &
-surfs['neutron shield NEtop SWbot'] &
+surfs['lower bound'] & -surfs['upper bound'])
root_univ.add_cell(cell)
cell = openmc.Cell(name='neutron shield panel SE')
cell.fill = mats['SS']
cell.region = (+surfs['core barrel OR'] & -surfs['neutron shield OR'] &
-surfs['neutron shield NWbot SEtop'] &
+surfs['neutron shield NWtop SEbot'] &
+surfs['lower bound'] & -surfs['upper bound'])
root_univ.add_cell(cell)
cell = openmc.Cell(name='neutron shield panel E')
cell.fill = mats['H2O']
cell.region = (+surfs['core barrel OR'] & -surfs['neutron shield OR'] &
+surfs['neutron shield NWbot SEtop'] &
+surfs['neutron shield NEbot SWtop'] &
+surfs['lower bound'] & -surfs['upper bound'])
root_univ.add_cell(cell)
cell = openmc.Cell(name='neutron shield panel NE')
cell.fill = mats['SS']
cell.region = (+surfs['core barrel OR'] & -surfs['neutron shield OR'] &
+surfs['neutron shield NEbot SWtop'] &
-surfs['neutron shield NEtop SWbot'] &
+surfs['lower bound'] & -surfs['upper bound'])
root_univ.add_cell(cell)
cell = openmc.Cell(name='neutron shield panel S')
cell.fill = mats['H2O']
cell.region = (+surfs['core barrel OR'] & -surfs['neutron shield OR'] &
-surfs['neutron shield NWtop SEbot'] &
+surfs['neutron shield NEtop SWbot'] &
+surfs['lower bound'] & -surfs['upper bound'])
root_univ.add_cell(cell)
cell = openmc.Cell(name='neutron shield panel SW')
cell.fill = mats['SS']
cell.region = (+surfs['core barrel OR'] & -surfs['neutron shield OR'] &
-surfs['neutron shield NEbot SWtop'] &
+surfs['neutron shield NEtop SWbot'] &
+surfs['lower bound'] & -surfs['upper bound'])
root_univ.add_cell(cell)
cell = openmc.Cell(name='neutron shield panel W')
cell.fill = mats['H2O']
cell.region = (+surfs['core barrel OR'] & -surfs['neutron shield OR'] &
-surfs['neutron shield NWbot SEtop'] &
-surfs['neutron shield NEbot SWtop'] &
+surfs['lower bound'] & -surfs['upper bound'])
root_univ.add_cell(cell)
#### CONSTRUCT DOWNCOMER
cell = openmc.Cell(name='downcomer')
cell.fill = mats['H2O']
cell.region = (+surfs['neutron shield OR'] & -surfs['RPV IR'] &
+surfs['lower bound'] & -surfs['upper bound'])
root_univ.add_cell(cell)
#### CONSTRUCT REACTOR PRESSURE VESSEL
cell = openmc.Cell(name='reactor pressure vessel')
cell.fill = mats['CS']
cell.region = (+surfs['RPV IR'] & -surfs['RPV OR'] &
+surfs['lower bound'] & -surfs['upper bound'])
root_univ.add_cell(cell)
#### CONSTRUCT GEOMETRY
geometry = openmc.Geometry(root_univ)
# Return geometry
return openmc.Geometry(root_univ)

139
smr/smr/materials.py
View file

@ -1,19 +1,55 @@
"""Instantiate the OpenMC Materials needed by the core model."""
import copy
import openmc
from openmc.data import atomic_weight, atomic_mass
from openmc.data import atomic_weight, atomic_mass, water_density
from . import system_pressure, core_average_temperature
_DEPLETION_NUCLIDES = [
"U239", "U240", "Np234", "Np235", "Np236", "Np237", "Np238", "Np239",
"Pu236", "Pu237", "Pu238", "Pu239", "Pu240", "Pu241", "Pu242", "B11",
"N14", "N15", "Fe57", "Fe58", "Co59", "Ni60", "Ni61", "Ni62",
"Cu63", "Ni64", "Zn64", "Cu65", "Zn65", "Zn66", "Zn67", "Zn68",
"Ga69", "Zn70", "Ge70", "Ga71", "Ge72", "Ge73", "Ge74", "As74",
"Se74", "As75", "Ge76", "Se76", "Se77", "Se78", "Se79", "Br79",
"Se80", "Kr80", "Br81", "Se82", "Kr82", "Kr83", "Kr84", "Sr84",
"Kr85", "Rb85", "Kr86", "Rb86", "Sr86", "Rb87", "Sr87", "Sr88",
"Sr89", "Y89", "Sr90", "Y90", "Zr90", "Y91", "Zr91", "Zr92",
"Zr93", "Nb93", "Zr94", "Nb94", "Mo94", "Zr95", "Nb95", "Mo95",
"Zr96", "Mo96", "Mo97", "Mo98", "Ru98", "Mo99", "Tc99", "Ru99",
"Mo100", "Ru100", "Ru101", "Ru102", "Pd102", "Ru103", "Rh103", "Ru104",
"Pd104", "Ru105", "Rh105", "Pd105", "Ru106", "Pd106", "Pd107", "Ag107",
"Pd108", "Cd108", "Ag109", "Pd110", "Cd110", "Ag111", "Cd111", "Cd112",
"Sn112", "Cd113", "In113", "Sn113", "Cd114", "Sn114", "In115", "Sn115",
"Cd116", "Sn116", "Sn117", "Sn118", "Sn119", "Sn120", "Te120", "Sb121",
"Sn122", "Te122", "Sn123", "Sb123", "Te123", "Sn124", "Sb124", "Te124",
"Sn125", "Sb125", "Te125", "Sn126", "Sb126", "Te126", "Xe126", "I127",
"Te128", "Xe128", "I129", "Xe129", "Te130", "I130", "Xe130", "I131",
"Xe131", "Te132", "Xe132", "Ba132", "Xe133", "Cs133", "Ba133", "Xe134",
"Cs134", "Ba134", "I135", "Xe135", "Cs135", "Ba135", "Xe136", "Cs136",
"Ba136", "Cs137", "Ba137", "Ba138", "La138", "Ce138", "La139", "Ce139",
"Ba140", "La140", "Ce140", "Ce141", "Pr141", "Ce142", "Pr142", "Nd142",
"Ce143", "Pr143", "Nd143", "Ce144", "Nd144", "Nd145", "Nd146", "Nd147",
"Pm147", "Sm147", "Nd148", "Pm148", "Sm148", "Pm149", "Sm149", "Nd150",
"Sm150", "Pm151", "Sm151", "Eu151", "Sm152", "Eu152", "Gd152", "Sm153",
"Eu153", "Gd153", "Sm154", "Eu154", "Gd154", "Eu155", "Gd155", "Eu156",
"Gd156", "Eu157", "Gd157", "Gd158", "Dy158", "Tb159", "Gd160", "Tb160",
"Dy160", "Dy161", "Dy162", "Dy163", "Dy164", "Er164", "Ho165", "Er166",
"Er167", "Er168", "Tm168", "Tm169", "Er170", "Tm170"]
mats = {}
# Create He gas material for fuel pin gap
mats['He'] = openmc.Material(name='Helium')
mats['He'].temperature = 300
mats['He'].set_density('g/cc', 0.0015981)
mats['He'].add_element('He', 1.0, 'ao')
# Create air material for instrument tubes
mats['Air'] = openmc.Material(name='Air')
mats['Air'].temperature = 300
mats['Air'].set_density('g/cc', 0.00616)
mats['Air'].add_element('O', 0.2095, 'ao')
mats['Air'].add_element('N', 0.7809, 'ao')
@ -22,7 +58,6 @@ mats['Air'].add_element('C', 0.00027, 'ao')
# Create inconel 718 material
mats['In'] = openmc.Material(name='Inconel')
mats['In'].temperature = 300
mats['In'].set_density('g/cc', 8.2)
mats['In'].add_element('Si', 0.0035, 'wo')
mats['In'].add_element('Cr', 0.1896, 'wo')
@ -30,9 +65,17 @@ mats['In'].add_element('Mn', 0.0087, 'wo')
mats['In'].add_element('Fe', 0.2863, 'wo')
mats['In'].add_element('Ni', 0.5119, 'wo')
# Create stainless steel 302
mats['SS302'] = openmc.Material(name='SS302')
mats['SS302'].set_density('g/cm3', 7.86)
mats['SS302'].add_element('Si', 0.01, 'wo')
mats['SS302'].add_element('Cr', 0.18, 'wo')
mats['SS302'].add_element('Mn', 0.02, 'wo')
mats['SS302'].add_element('Fe', 0.70, 'wo')
mats['SS302'].add_element('Ni', 0.09, 'wo')
# Create stainless steel material
mats['SS'] = openmc.Material(name='SS304')
mats['SS'].temperature = 300
mats['SS'].set_density('g/cc', 8.03)
mats['SS'].add_element('Si', 0.0060, 'wo')
mats['SS'].add_element('Cr', 0.1900, 'wo')
@ -42,7 +85,6 @@ mats['SS'].add_element('Ni', 0.1000, 'wo')
# Create carbon steel material
mats['CS'] = openmc.Material(name='Carbon Steel')
mats['CS'].temperature = 300
mats['CS'].set_density('g/cc', 7.8)
mats['CS'].add_element('C', 0.00270, 'wo')
mats['CS'].add_element('Mn', 0.00750, 'wo')
@ -63,7 +105,6 @@ mats['CS'].add_element('Fe', 0.96487, 'wo')
# Create zircaloy 4 material
mats['Zr'] = openmc.Material(name='Zircaloy-4')
mats['Zr'].temperature = 300
mats['Zr'].set_density('g/cc', 6.55)
mats['Zr'].add_element('O', 0.00125, 'wo')
mats['Zr'].add_element('Cr', 0.0010, 'wo')
@ -71,9 +112,18 @@ mats['Zr'].add_element('Fe', 0.0021, 'wo')
mats['Zr'].add_element('Zr', 0.98115, 'wo')
mats['Zr'].add_element('Sn', 0.0145, 'wo')
# Create M5 alloy material
m5_niobium = 0.01 # http://publications.jrc.ec.europa.eu/repository/bitstream/JRC100644/lcna28366enn.pdf
m5_oxygen = 0.00135 # http://publications.jrc.ec.europa.eu/repository/bitstream/JRC100644/lcna28366enn.pdf
m5_density = 6.494 # 10.1039/C5DT03403E
mats['M5'] = openmc.Material(name='M5')
mats['M5'].add_element('Zr', 1.0 - m5_niobium - m5_oxygen)
mats['M5'].add_element('Nb', m5_niobium)
mats['M5'].add_element('O', m5_oxygen)
mats['M5'].set_density('g/cm3', m5_density)
# Create Ag-In-Cd control rod material
mats['AIC'] = openmc.Material(name='Ag-In-Cd')
mats['AIC'].temperature = 300
mats['AIC'].set_density('g/cc', 10.16)
mats['AIC'].add_element('Ag', 0.80, 'wo')
mats['AIC'].add_element('In', 0.15, 'wo')
@ -82,10 +132,11 @@ mats['AIC'].add_element('Cd', 0.05, 'wo')
#### Borated Water
boron_ppm = 975
# Concentration of boron at beginning of equilibrium cycle
boron_ppm = 1240 # ML17013A274, Figure 4.3-17
# Density of clean water at 2250 psia T=560F NIST
h2o_dens = 0.73986
# Density of water
h2o_dens = water_density(core_average_temperature, system_pressure)
# Weight percent of natural boron in borated water
wB_Bh2o = boron_ppm * 1.0e-6
@ -112,7 +163,6 @@ aho_Bh2o = ah2o_Bh2o
# Create borated water for coolant / moderator
mats['H2O'] = openmc.Material(name='Borated Water')
mats['H2O'].temperature = 300
mats['H2O'].set_density('g/cc', rho_Bh2o)
mats['H2O'].add_element('B', aB_Bh2o, 'ao')
mats['H2O'].add_element('H', ah_Bh2o, 'ao')
@ -144,7 +194,6 @@ aB_bsg = aB10_bsg + aB11_bsg
# Create borosilicate glass material
mats['BSG'] = openmc.Material(name='Borosilicate Glass')
mats['BSG'].temperature = 300
mats['BSG'].set_density('g/cc', 2.26)
mats['BSG'].add_element('O', aO_bsg, 'ao')
mats['BSG'].add_element('Si', aSi_bsg, 'ao')
@ -156,27 +205,59 @@ mats['BSG'].add_nuclide('B11', aB11_bsg, 'ao')
#### Enriched UO2 Fuel
# Create 1.6% enriched UO2 fuel material
mats['UO2 1.6'] = openmc.Material(name='1.6% Enr. UO2 Fuel')
mats['UO2 1.6'].temperature = 300
mats['UO2 1.6'].set_density('g/cc', 10.31341)
mats['UO2 1.6'].add_element('O', 2., 'ao')
mats['UO2 1.6'].add_element('U', 1., 'ao', enrichment=1.61006)
mat = openmc.Material(name='1.6% Enr. UO2 Fuel')
mat.set_density('g/cc', 10.31341)
mat.add_element('O', 2., 'ao')
mat.add_element('U', 1., 'ao', enrichment=1.61006)
mats['UO2 1.6 fresh'] = mat
# Create 2.4% enriched UO2 fuel material
mats['UO2 2.4'] = openmc.Material(name='2.4% Enr. UO2 Fuel')
mats['UO2 2.4'].temperature = 300
mats['UO2 2.4'].set_density('g/cc', 10.29748)
mats['UO2 2.4'].add_element('O', 2., 'ao')
mats['UO2 2.4'].add_element('U', 1., 'ao', enrichment=2.39993)
mat = openmc.Material(name='2.4% Enr. UO2 Fuel')
mat.set_density('g/cc', 10.29748)
mat.add_element('O', 2., 'ao')
mat.add_element('U', 1., 'ao', enrichment=2.39993)
mats['UO2 2.4 fresh'] = mat
# Create 3.1% enriched UO2 fuel material
mats['UO2 3.1'] = openmc.Material(name='3.1% Enr. UO2 Fuel')
mats['UO2 3.1'].temperature = 300
mats['UO2 3.1'].set_density('g/cc', 10.30166)
mats['UO2 3.1'].add_element('O', 2., 'ao')
mats['UO2 3.1'].add_element('U', 1., 'ao', enrichment=3.10221)
mat = openmc.Material(name='3.1% Enr. UO2 Fuel')
mat.set_density('g/cc', 10.30166)
mat.add_element('O', 2., 'ao')
mat.add_element('U', 1., 'ao', enrichment=3.10221)
mats['UO2 3.1 fresh'] = mat
# Depleted versions of 1.6%, 2.4%, 3.1% fuel
mat = openmc.Material(name='2.4% Enr. UO2 Fuel')
mat.set_density('g/cc', 10.29748)
mat.add_element('O', 2., 'ao')
mat.add_element('U', 1., 'ao', enrichment=2.39993)
for nuc in _DEPLETION_NUCLIDES:
mat.add_nuclide(nuc, 1.0e-11)
mats['UO2 2.4 depleted'] = mat
mat = openmc.Material(name='1.6% Enr. UO2 Fuel')
mat.set_density('g/cc', 10.31341)
mat.add_element('O', 2., 'ao')
mat.add_element('U', 1., 'ao', enrichment=1.61006)
for nuc in _DEPLETION_NUCLIDES:
mat.add_nuclide(nuc, 1.0e-11)
mats['UO2 1.6 depleted'] = mat
mat = openmc.Material(name='3.1% Enr. UO2 Fuel')
mat.set_density('g/cc', 10.30166)
mat.add_element('O', 2., 'ao')
mat.add_element('U', 1., 'ao', enrichment=3.10221)
for nuc in _DEPLETION_NUCLIDES:
mat.add_nuclide(nuc, 1.0e-11)
mats['UO2 3.1 depleted'] = mat
# Construct a collection of Materials to export to XML
materials = openmc.Materials(mats.values())
materials = openmc.Materials(mats.values())
def clone(material):
"""Perform copy of material but share nuclide densities"""
shared_mat = copy.copy(material)
shared_mat.id = None
return shared_mat

1356
smr/smr/pins.py
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File diff suppressed because it is too large Load diff

175
smr/smr/plots.py
View file

@ -1,69 +1,134 @@
"""Instantiate OpenMC Plots to visualize the core model."""
from random import randint, seed
import openmc
from .surfaces import lowest_extent, highest_extent, lattice_pitch, rpv_OR
from .surfaces import lowest_extent, highest_extent, lattice_pitch, rpv_OR, \
pin_pitch
from .materials import mats
# color specifications
colors = {mats['H2O']: [198, 226, 255], # light blue
mats['In']: [101, 101, 101], # dgray
mats['CS']: [ 0, 0, 0], # carbons black
mats['Zr']: [201, 201, 201], # gray
mats['SS']: [ 0, 0, 0], # black
mats['Air']: [255, 255, 255], # white
mats['He']: [255, 218, 185], # light orange
mats['BSG']: [ 0, 255, 0], # green
mats['AIC']: [255, 0, 0], # bright red
mats['UO2 1.6']: [142, 35, 35], # light red
mats['UO2 2.4']: [255, 215, 0], # gold
mats['UO2 3.1']: [ 0, 0, 128]} # dark blue
colors = {
mats['H2O']: (198, 226, 255), # light blue
mats['In']: (101, 101, 101), # dgray
mats['CS']: ( 0, 0, 0), # carbons black
mats['Zr']: (201, 201, 201), # gray
mats['SS']: ( 0, 0, 0), # black
mats['Air']: (255, 255, 255), # white
mats['He']: (255, 218, 185), # light orange
mats['BSG']: ( 0, 255, 0), # green
mats['AIC']: (255, 0, 0), # bright red
mats['UO2 1.6 fresh']: (142, 35, 35), # light red
mats['UO2 2.4 fresh']: (255, 215, 0), # gold
mats['UO2 3.1 fresh']: ( 0, 0, 128) # dark blue
}
# Create a collection of plots
plots = openmc.Plots()
def core_plots():
# Create a collection of plots
plots = openmc.Plots()
plot = openmc.Plot(name='radial slice ')
plot.basis = 'xy'
plot.color_by = 'material'
plot.origin = [0., 0., (highest_extent-lowest_extent)/2.]
plot.width = [25*lattice_pitch/2, 25*lattice_pitch/2.]
plot.filename = 'radial_xy_slice'
plot.colors = colors
plot.background = [255, 255, 255]
plot.pixels = [1000, 1000]
plots += [plot]
plot = openmc.Plot(name='radial slice ')
plot.basis = 'xy'
plot.color_by = 'material'
plot.origin = [0., 0., (highest_extent-lowest_extent)/2.]
plot.width = [25*lattice_pitch/2, 25*lattice_pitch/2.]
plot.filename = 'radial_xy_slice'
plot.colors = colors
plot.background = [255, 255, 255]
plot.pixels = [1000, 1000]
plots.append(plot)
plot = openmc.Plot(name='axial slice')
plot.basis = 'xz'
plot.color_by = 'material'
plot.origin = [0., 0., (highest_extent-lowest_extent)/2.]
plot.width = [rpv_OR*2., (highest_extent-lowest_extent)]
plot.filename = 'axial_xz_slice'
plot.colors = colors
plot.background = [255, 255, 255]
plot.pixels = [1000, 1000]
plots += [plot]
plot = openmc.Plot(name='axial slice')
plot.basis = 'xz'
plot.color_by = 'material'
plot.origin = [0., 0., (highest_extent-lowest_extent)/2.]
plot.width = [rpv_OR*2., (highest_extent-lowest_extent)]
plot.filename = 'axial_xz_slice'
plot.colors = colors
plot.background = [255, 255, 255]
plot.pixels = [1000, 1000]
plots.append(plot)
plot = openmc.Plot(name='assembly grid spacer')
plot.basis = 'xy'
plot.color_by = 'material'
plot.origin = [0., 0., 95.0]
plot.width = [lattice_pitch*1.5, lattice_pitch*1.5]
plot.filename = 'assm_grid_spacer'
plot.colors = colors
plot.background = [255, 255, 255]
plot.pixels = [2000, 2000]
plots += [plot]
plot = openmc.Plot(name='assembly grid spacer')
plot.basis = 'xy'
plot.color_by = 'material'
plot.origin = [0., 0., 95.0]
plot.width = [lattice_pitch*1.5, lattice_pitch*1.5]
plot.filename = 'assm_grid_spacer'
plot.colors = colors
plot.background = [255, 255, 255]
plot.pixels = [2000, 2000]
plots.append(plot)
plot = openmc.Plot(name='assembly no spacer')
plot.basis = 'xy'
plot.color_by = 'material'
plot.origin = [0., 0., 90.]
plot.width = [lattice_pitch*1.5, lattice_pitch*1.5]
plot.filename = 'assm_no_spacer'
plot.colors = colors
plot.background = [255, 255, 255]
plot.pixels = [2000, 2000]
plots += [plot]
plot = openmc.Plot(name='assembly no spacer')
plot.basis = 'xy'
plot.color_by = 'material'
plot.origin = [0., 0., 90.]
plot.width = [lattice_pitch*1.5, lattice_pitch*1.5]
plot.filename = 'assm_no_spacer'
plot.colors = colors
plot.background = [255, 255, 255]
plot.pixels = [2000, 2000]
plots.append(plot)
plot = openmc.Plot(name='assembly no spacer cell')
plot.basis = 'xy'
plot.color_by = 'cell'
plot.origin = [0., 0., 90.]
plot.width = [lattice_pitch*1.5, lattice_pitch*1.5]
plot.filename = 'assm_no_spacer_cell'
plot.background = [255, 255, 255]
plot.pixels = [2000, 2000]
plots.append(plot)
plot = openmc.Plot(name='z slice')
plot.basis = 'xz'
plot.color_by = 'cell'
plot.filename = 'assm_xz'
plot.origin = (0., 0., lowest_extent + (highest_extent - lowest_extent)/2)
plot.width = (lattice_pitch*1.5, highest_extent - lowest_extent)
plot.pixels = (int(plot.width[0]/plot.width[1]*4000), 4000)
plots.append(plot)
return plots
def assembly_plots(univ):
colors = {
mats['H2O']: (198, 226, 255), # light blue
mats['In']: (101, 101, 101), # dgray
mats['Zr']: (201, 201, 201), # gray
mats['SS']: ( 0, 0, 0), # black
mats['Air']: (255, 255, 255), # white
mats['He']: (255, 218, 185), # light orange
mats['BSG']: ( 0, 255, 0), # green
}
# Set colors for fuel as shades of yellow
seed(1)
for m in univ.get_all_materials().values():
if 'UO2 Fuel' in m.name:
colors[m] = (randint(200, 255), randint(200, 255), randint(0, 150))
pin = openmc.Plot(name='fuel pin')
pin.basis = 'xy'
pin.color_by = 'material'
pin.colors = colors
pin.filename = 'smr_pin'
pin.origin = (pin_pitch, 0., (highest_extent-lowest_extent)/2.)
pin.width = (pin_pitch, pin_pitch)
pin.pixels = (400, 400)
assm = openmc.Plot(name='fuel assembly')
assm.basis = 'xy'
assm.color_by = 'material'
assm.colors = colors
assm.filename = 'smr_assm'
assm.origin = (0., 0., (highest_extent-lowest_extent)/2.)
assm.width = (lattice_pitch, lattice_pitch)
assm.pixels = (2000, 2000)
return openmc.Plots([pin, assm])

397
smr/smr/reflector.py
View file

@ -12,8 +12,7 @@ converted to actual dimensions by scaling according to the width of an assembly.
import openmc
from .materials import mats
from .surfaces import surfs, lattice_pitch
from .assemblies import univs
from smr import surfaces
def make_reflector(name, parameters):
@ -27,285 +26,205 @@ def make_reflector(name, parameters):
Iterable containing tuple with the (x,y) coordinates of the center and
the radius of a Z-cylinder and the
Returns
-------
openmc.Universe
Universe containing reflector block
"""
water_holes = []
for x, y, r in parameters:
zcyl = openmc.ZCylinder(x0=x, y0=y, R=r)
zcyl = openmc.ZCylinder(x0=x, y0=y, r=r)
hole = openmc.Cell(fill=mats['H2O'], region=-zcyl)
water_holes.append(hole)
ss_region = openmc.Intersection(~c.region for c in water_holes)
ss_cell = openmc.Cell(name='{} SS'.format(name), fill=mats['SS'],
ss_cell = openmc.Cell(name='reflector {} SS'.format(name), fill=mats['SS'],
region=ss_region)
univs[name] = openmc.Universe(name=name)
univs[name].add_cells(water_holes)
univs[name].add_cell(ss_cell)
univ = openmc.Universe(name='reflector {}'.format(name))
univ.add_cells(water_holes)
univ.add_cell(ss_cell)
return univ
# Reflector at northwest corner (fuel assemblies to the right and below)
def reflector_universes():
"""Generate universes for SMR heavy neutron reflector blocks.
width = 276
p1 = 59
p2 = 126
p3 = 196
p4 = 264
Parameters
----------
num_rings : int
Number of annual regions in fuel
num_axial : int
Number of axial subdivisions in fuel
p5 = 105
Returns
-------
dict
Dictionary mapping a universe name to a openmc.Universe object
p6 = 122
p7 = 164
"""
# Create dictionary to store universes
univs = {}
p8 = 138
p9 = 222
# Reflector at northwest corner (fuel assemblies to the right and below)
width = 276
p1 = 59
p2 = 126
p3 = 196
p4 = 264
p10 = 247
p5 = 105
# There are 8 large water holes and all others appear to have the same, smaller
# diameter
d_small = 13
d_large = 30
p6 = 122
p7 = 164
# All pixel widths are scaled according to the actual width of an assembly
# divided by the width of an assembly in pixels
scale = lattice_pitch/width
p8 = 138
p9 = 222
# Physical positions
x1 = -lattice_pitch/2 + scale*(width - p4)
x2 = -lattice_pitch/2 + scale*(width - p3)
x3 = -lattice_pitch/2 + scale*(width - p2)
x4 = -lattice_pitch/2 + scale*(width - p1)
y1 = -lattice_pitch/2 + scale*p1
y2 = -lattice_pitch/2 + scale*p2
y3 = -lattice_pitch/2 + scale*p3
y4 = -lattice_pitch/2 + scale*p4
p10 = 247
x5 = -lattice_pitch/2 + scale*(width - p5)
y5 = -lattice_pitch/2 + scale*p5
x6 = -lattice_pitch/2 + scale*(width - p7)
y6 = -lattice_pitch/2 + scale*p6
x7 = -lattice_pitch/2 + scale*(width - p6)
y7 = -lattice_pitch/2 + scale*p7
x8 = -lattice_pitch/2 + scale*(width - p9)
y8 = -lattice_pitch/2 + scale*p8
x9 = -lattice_pitch/2 + scale*(width - p8)
y9 = -lattice_pitch/2 + scale*p9
# There are 8 large water holes and all others appear to have the same, smaller
# diameter
d_small = 13
d_large = 30
y10 = -lattice_pitch/2 + scale*p10
# All pixel widths are scaled according to the actual width of an assembly
# divided by the width of an assembly in pixels
lattice_pitch = surfaces.lattice_pitch
scale = lattice_pitch/width
# Radius of small/large water holes
r1 = scale*d_small/2
r2 = scale*d_large/2
# Physical positions
x1 = -lattice_pitch/2 + scale*(width - p4)
x2 = -lattice_pitch/2 + scale*(width - p3)
x3 = -lattice_pitch/2 + scale*(width - p2)
x4 = -lattice_pitch/2 + scale*(width - p1)
y1 = -lattice_pitch/2 + scale*p1
y2 = -lattice_pitch/2 + scale*p2
y3 = -lattice_pitch/2 + scale*p3
y4 = -lattice_pitch/2 + scale*p4
params = [
(x1, y1, r1), (x2, y1, r1), (x3, y1, r1), (x4, y1, r2),
(x4, y2, r1), (x4, y3, r1), (x4, y4, r1), (x5, y5, r1),
(x6, y6, r1), (x7, y7, r1), (x8, y8, r1), (x9, y9, r1),
(x1, y10, r1)
]
x5 = -lattice_pitch/2 + scale*(width - p5)
y5 = -lattice_pitch/2 + scale*p5
x6 = -lattice_pitch/2 + scale*(width - p7)
y6 = -lattice_pitch/2 + scale*p6
x7 = -lattice_pitch/2 + scale*(width - p6)
y7 = -lattice_pitch/2 + scale*p7
x8 = -lattice_pitch/2 + scale*(width - p9)
y8 = -lattice_pitch/2 + scale*p8
x9 = -lattice_pitch/2 + scale*(width - p8)
y9 = -lattice_pitch/2 + scale*p9
make_reflector('heavy reflector NW', params)
y10 = -lattice_pitch/2 + scale*p10
# Reflector at (1, 1)
# Radius of small/large water holes
r1 = scale*d_small/2
r2 = scale*d_large/2
params = [
(x4, y1, r1),
(lattice_pitch/2 - scale*103, -lattice_pitch/2 + scale*156, r1),
(lattice_pitch/2 - scale*158, -lattice_pitch/2 + scale*103, r1)
]
make_reflector('heavy reflector 1,1', params)
params = [
(x1, y1, r1), (x2, y1, r1), (x3, y1, r1), (x4, y1, r2),
(x4, y2, r1), (x4, y3, r1), (x4, y4, r1), (x5, y5, r1),
(x6, y6, r1), (x7, y7, r1), (x8, y8, r1), (x9, y9, r1),
(x1, y10, r1)
]
univs['NW'] = make_reflector('NW', params)
# Left reflector (4,0)
# Reflector at (1, 1)
left1 = 58
left2 = 118
left3 = 173
up3 = 76
params = [
(x4, y1, r1),
(lattice_pitch/2 - scale*103, -lattice_pitch/2 + scale*156, r1),
(lattice_pitch/2 - scale*158, -lattice_pitch/2 + scale*103, r1)
]
univs['1,1'] = make_reflector('1,1', params)
x1 = -lattice_pitch/2 + scale*(width - left1)
x2 = -lattice_pitch/2 + scale*(width - left2)
d_y = scale*67
x3 = -lattice_pitch/2 + scale*(width - left3)
y3 = scale*up3
# Left reflector (4,0)
params = [
(x1, 0, r1), (x1, d_y, r1), (x1, 2*d_y, r1), (x1, -d_y, r1), (x1, -2*d_y, r1),
(x2, d_y/2, r1), (x2, 3/2*d_y, r1), (x2, -d_y/2, r1), (x2, -3/2*d_y, r1),
(x3, y3, r1), (x3, -y3, r1)
]
left1 = 58
left2 = 118
left3 = 173
up3 = 76
make_reflector('heavy reflector 4,0', params)
x1 = -lattice_pitch/2 + scale*(width - left1)
x2 = -lattice_pitch/2 + scale*(width - left2)
d_y = scale*67
x3 = -lattice_pitch/2 + scale*(width - left3)
y3 = scale*up3
# Reflector at (3,0)
params = [
(x1, 0, r1), (x1, d_y, r1), (x1, 2*d_y, r1), (x1, -d_y, r1), (x1, -2*d_y, r1),
(x2, d_y/2, r1), (x2, 3/2*d_y, r1), (x2, -d_y/2, r1), (x2, -3/2*d_y, r1),
(x3, y3, r1), (x3, -y3, r1)
]
univs['4,0'] = make_reflector('4,0', params)
params = []
for i in range(2, 7):
params.append((x1, i*d_y - lattice_pitch, r1))
for i in (5, 7, 11):
params.append((x2, i*d_y/2 - lattice_pitch, r1))
# Reflector at (3,0)
left3 = 140
left4 = 183
up3 = 159
up4 = 47
params = []
for i in range(2, 7):
params.append((x1, i*d_y - lattice_pitch, r1))
for i in (5, 7, 11):
params.append((x2, i*d_y/2 - lattice_pitch, r1))
x3 = -lattice_pitch/2 + scale*(width - left3)
y3 = -lattice_pitch/2 + scale*up3
x4 = -lattice_pitch/2 + scale*(width - left4)
y4 = -lattice_pitch/2 + scale*up4
params += [(x3, y3, r1), (x4, y4, r1)]
left3 = 140
left4 = 183
up3 = 159
up4 = 47
make_reflector('heavy reflector 3,0', params)
x3 = -lattice_pitch/2 + scale*(width - left3)
y3 = -lattice_pitch/2 + scale*up3
x4 = -lattice_pitch/2 + scale*(width - left4)
y4 = -lattice_pitch/2 + scale*up4
params += [(x3, y3, r1), (x4, y4, r1)]
# Reflector at (5,0)
univs['3,0'] = make_reflector('3,0', params)
params = [(x, -y, r) for x, y, r in params]
make_reflector('heavy reflector 5,0', params)
# Reflector at (5,0)
params = [(x, -y, r) for x, y, r in params]
univs['5,0'] = make_reflector('5,0', params)
# Reflector at (2, 0)
# Reflector at (2, 0)
params = [(-lattice_pitch/2 + scale*(width - 78),
-lattice_pitch/2 + scale*98, r1)]
make_reflector('heavy reflector 2,0', params)
params = [(-lattice_pitch/2 + scale*(width - 78),
-lattice_pitch/2 + scale*98, r1)]
univs['2,0'] = make_reflector('2,0', params)
################################################################################
# Beyond this point, all universes are just copies of the ones previously
# created with a rotation applied
################################################################################
# Beyond this point, all universes are just copies of the ones previously
# created with a rotation applied
# NE corner
cell = openmc.Cell(name='heavy reflector NE', fill=univs['heavy reflector NW'])
cell.rotation = (0, 0, -90)
univs['heavy reflector NE'] = openmc.Universe(name='heavy reflector NE')
univs['heavy reflector NE'].add_cell(cell)
# First define helper function to create new universe by rotating an
# existing one
def rotate_universe(univ, rotation, name):
cell = openmc.Cell(name='reflector {}'.format(name), fill=univ)
cell.rotation = rotation
return openmc.Universe(name=name, cells=[cell])
# SW corner
cell = openmc.Cell(name='heavy reflector SW', fill=univs['heavy reflector NW'])
cell.rotation = (0, 0, 90)
univs['heavy reflector SW'] = openmc.Universe(name='heavy reflector SW')
univs['heavy reflector SW'].add_cell(cell)
univs['NE'] = rotate_universe(univs['NW'], (0, 0, -90), 'NE')
univs['SW'] = rotate_universe(univs['NW'], (0, 0, 90), 'SW')
univs['SE'] = rotate_universe(univs['NW'], (0, 0, 180), 'SE')
univs['0,2'] = rotate_universe(univs['2,0'], (0, 180, -90), '0,2')
univs['0,3'] = rotate_universe(univs['5,0'], (0, 0, -90), '0,3')
univs['0,4'] = rotate_universe(univs['4,0'], (0, 0, -90), '0,4')
univs['0,5'] = rotate_universe(univs['3,0'], (0, 0, -90), '0,5')
univs['0,6'] = rotate_universe(univs['2,0'], (0, 0, -90), '0,6')
univs['1,7'] = rotate_universe(univs['1,1'], (0, 0, -90), '1,7')
univs['2,8'] = rotate_universe(univs['2,0'], (0, 180, 0), '2,8')
univs['3,8'] = rotate_universe(univs['3,0'], (0, 180, 0), '3,8')
univs['4,8'] = rotate_universe(univs['4,0'], (0, 180, 0), '4,8')
univs['5,8'] = rotate_universe(univs['3,0'], (0, 0, 180), '5,8')
univs['6,0'] = rotate_universe(univs['2,0'], (180, 0, 0), '6,0')
univs['6,8'] = rotate_universe(univs['2,0'], (0, 0, 180), '6,8')
univs['7,1'] = rotate_universe(univs['1,1'], (180, 0, 0), '7,1')
univs['7,7'] = rotate_universe(univs['1,1'], (0, 0, 180), '7,7')
univs['8,2'] = rotate_universe(univs['2,0'], (0, 0, 90), '8,2')
univs['8,3'] = rotate_universe(univs['3,0'], (0, 0, 90), '8,3')
univs['8,4'] = rotate_universe(univs['4,0'], (0, 0, 90), '8,4')
univs['8,5'] = rotate_universe(univs['5,0'], (0, 0, 90), '8,5')
univs['8,6'] = rotate_universe(univs['2,0'], (0, 0, 180), '8,6')
# SE corner
cell = openmc.Cell(name='heavy reflector SE', fill=univs['heavy reflector NW'])
cell.rotation = (0, 0, 180)
univs['heavy reflector SE'] = openmc.Universe(name='heavy reflector SE')
univs['heavy reflector SE'].add_cell(cell)
# Solid stainless steel universe
all_ss = openmc.Cell(name='heavy reflector', fill=mats['SS'])
univs['solid'] = openmc.Universe(name='solid', cells=[all_ss])
# Reflector at (0, 2)
name = 'heavy reflector 0,2'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 2,0'])
cell.rotation = (0, 180, -90)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (0, 3)
name = 'heavy reflector 0,3'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 5,0'])
cell.rotation = (0, 0, -90)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (0, 4)
name = 'heavy reflector 0,4'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 4,0'])
cell.rotation = (0, 0, -90)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (0, 5)
name = 'heavy reflector 0,5'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 3,0'])
cell.rotation = (0, 0, -90)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (0, 6)
name = 'heavy reflector 0,6'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 2,0'])
cell.rotation = (0, 0, -90)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (1, 7)
name = 'heavy reflector 1,7'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 1,1'])
cell.rotation = (0, 0, -90)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (2, 8)
name = 'heavy reflector 2,8'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 2,0'])
cell.rotation = (0, 180, 0)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (3, 8)
name = 'heavy reflector 3,8'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 3,0'])
cell.rotation = (0, 180, 0)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (4, 8)
name = 'heavy reflector 4,8'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 4,0'])
cell.rotation = (0, 180, 0)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (5, 8)
name = 'heavy reflector 5,8'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 3,0'])
cell.rotation = (0, 0, 180)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (6, 0)
name = 'heavy reflector 6,0'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 2,0'])
cell.rotation = (180, 0, 0)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (6, 8)
name = 'heavy reflector 6,8'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 2,0'])
cell.rotation = (0, 0, 180)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (7, 1)
name = 'heavy reflector 7,1'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 1,1'])
cell.rotation = (180, 0, 0)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (7, 7)
name = 'heavy reflector 7,7'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 1,1'])
cell.rotation = (0, 0, 180)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (8, 2)
name = 'heavy reflector 8,2'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 2,0'])
cell.rotation = (0, 0, 90)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (8, 3)
name = 'heavy reflector 8,3'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 3,0'])
cell.rotation = (0, 0, 90)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (8, 4)
name = 'heavy reflector 8,4'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 4,0'])
cell.rotation = (0, 0, 90)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (8, 5)
name = 'heavy reflector 8,5'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 5,0'])
cell.rotation = (0, 0, 90)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Reflector at (8, 6)
name = 'heavy reflector 8,6'
cell = openmc.Cell(name=name, fill=univs['heavy reflector 2,0'])
cell.rotation = (0, 0, 180)
univs[name] = openmc.Universe(name=name, cells=[cell])
# Solid stainless steel universe
all_ss = openmc.Cell(name='heavy reflector', fill=mats['SS'])
univs['heavy reflector'] = openmc.Universe(
name='heavy reflector', cells=[all_ss])
return univs

180
smr/smr/surfaces.py
View file

@ -13,7 +13,6 @@ NuScale DC application, chapter 4: https://www.nrc.gov/docs/ML1701/ML17013A274.p
"""
import copy
from math import tan, pi
import numpy as np
@ -34,12 +33,13 @@ INCHES = 2.54
# fuel rod parameters
pellet_OR = 0.3195*INCHES/2 # ML17013A274, Table 4.1-2
pellet_length = 0.4*INCHES # ML17013A274, Table 4.1-2
clad_IR = 0.326*INCHES/2 # ML17013A274, Table 4.1-2
clad_OR = 0.374*INCHES/2 # ML17013A274, Table 4.1-2
active_fuel_length = 78.74*INCHES # ML17013A274, Figure 4.2-10
plenum_length = 5.311*INCHES # ML17013A274, Figure 4.2-10
fuel_rod_length = 85.00*INCHES # ML17013A274, Table 4.1-2
lower_end_cap = 0.575*INCHES # ML17007A001, Table 3-2
lower_end_cap_length = 0.575*INCHES # ML17007A001, Table 3-2
# pin cell parameters
guide_tube_IR = 0.450*INCHES/2 # ML17013A274, Table 4.1-2
@ -78,19 +78,20 @@ top_nozzle_width = 8.406*INCHES # ML17013A274, Figure 4.2-2
core_barrel_IR = 74*INCHES/2 # ML17013A274, Table 4.1-2
core_barrel_OR = 78*INCHES/2 # ML17013A274, Table 4.1-2
neutron_shield_OR = core_barrel_OR + 2.0
rpv_IR = 120.0 # Estimate?
rpv_OR = 135.0 # Estimate?
rpv_IR = 96.5*INCHES/2 # ML17013A274, Table 5.3-1
rpv_OR = 105*INCHES/2 # ML17013A274, Table 5.3-1
# axial parameters
lowest_extent = 0.000
bottom_support_plate = 20.000
top_support_plate = 25.000
bottom_lower_nozzle = 25.000
top_lower_nozzle = 35.160
bottom_fuel_rod = 35.160
top_lower_thimble = 36.007
bottom_fuel_stack = 36.007
bot_burn_abs = 41.087
reference_z = -36.6205
lowest_extent = reference_z
bottom_support_plate = lowest_extent + 20.000
top_support_plate = bottom_support_plate + 5.000
bottom_lower_nozzle = bottom_support_plate + 5.000
top_lower_nozzle = bottom_lower_nozzle + 4.0*INCHES
bottom_fuel_rod = bottom_lower_nozzle + 4.0*INCHES
top_lower_thimble = bottom_fuel_rod + lower_end_cap_length
bottom_fuel_stack = bottom_fuel_rod + lower_end_cap_length
bot_burn_abs = bottom_fuel_stack + 2.0*INCHES
top_active_core = bottom_fuel_stack + active_fuel_length
top_plenum = top_active_core + plenum_length
top_fuel_rod = bottom_fuel_rod + fuel_rod_length
@ -126,74 +127,47 @@ neutron_shield_NEtop_SWbot = tan(-pi/6)
surfs = {}
surfs['pellet OR'] = openmc.ZCylinder(
R=pellet_OR, name='Pellet OR')
surfs['plenum spring OR'] = openmc.ZCylinder(
R=plenum_spring_OR, name='FR Plenum Spring OR')
surfs['clad IR'] = openmc.ZCylinder(
R=clad_IR, name='Clad IR')
surfs['clad OR'] = openmc.ZCylinder(
R=clad_OR, name='Clad OR')
surfs['GT IR'] = openmc.ZCylinder(
R=guide_tube_IR, name='GT IR (above dashpot)')
surfs['GT OR'] = openmc.ZCylinder(
R=guide_tube_OR, name='GT OR (above dashpot)')
surfs['GT dashpot IR'] = openmc.ZCylinder(
R=guide_tube_dash_IR, name='GT IR (at dashpot)')
surfs['GT dashpot OR'] = openmc.ZCylinder(
R=guide_tube_dash_OR, name='GT OR (at dashpot)')
surfs['CP OR'] = openmc.ZCylinder(
R=boron_carbide_OR, name='Control Poison OR')
surfs['CR IR'] = openmc.ZCylinder(
R=control_rod_IR, name='CR Clad IR')
surfs['CR OR'] = openmc.ZCylinder(
R=control_rod_OR, name='CR Clad OR')
surfs['BA IR 1'] = openmc.ZCylinder(
R=burn_abs_r1, name='BA IR 1')
surfs['BA IR 2'] = openmc.ZCylinder(
R=burn_abs_r2, name='BA IR 2')
surfs['BA IR 3'] = openmc.ZCylinder(
R=burn_abs_r3, name='BA IR 3')
surfs['BA IR 4'] = openmc.ZCylinder(
R=burn_abs_r4, name='BA IR 4')
surfs['BA IR 5'] = openmc.ZCylinder(
R=burn_abs_r5, name='BA IR 5')
surfs['BA IR 6'] = openmc.ZCylinder(
R=burn_abs_r6, name='BA IR 6')
surfs['BA IR 7'] = openmc.ZCylinder(
R=burn_abs_r7, name='BA IR 7')
surfs['BA IR 8'] = openmc.ZCylinder(
R=burn_abs_r8, name='BA IR 8')
surfs['IT IR'] = copy.deepcopy(surfs['BA IR 5'])
surfs['IT OR'] = copy.deepcopy(surfs['BA IR 6'])
surfs['pellet OR'] = openmc.ZCylinder(r=pellet_OR, name='Pellet OR')
surfs['plenum spring OR'] = openmc.ZCylinder(r=plenum_spring_OR, name='FR Plenum Spring OR')
surfs['clad IR'] = openmc.ZCylinder(r=clad_IR, name='Clad IR')
surfs['clad OR'] = openmc.ZCylinder(r=clad_OR, name='Clad OR')
surfs['GT IR'] = openmc.ZCylinder(r=guide_tube_IR, name='GT IR (above dashpot)')
surfs['GT OR'] = openmc.ZCylinder(r=guide_tube_OR, name='GT OR (above dashpot)')
surfs['GT dashpot IR'] = openmc.ZCylinder(r=guide_tube_dash_IR, name='GT IR (at dashpot)')
surfs['GT dashpot OR'] = openmc.ZCylinder(r=guide_tube_dash_OR, name='GT OR (at dashpot)')
surfs['CP OR'] = openmc.ZCylinder(r=boron_carbide_OR, name='Control Poison OR')
surfs['CR IR'] = openmc.ZCylinder(r=control_rod_IR, name='CR Clad IR')
surfs['CR OR'] = openmc.ZCylinder(r=control_rod_OR, name='CR Clad OR')
surfs['BA IR 1'] = openmc.ZCylinder(r=burn_abs_r1, name='BA IR 1')
surfs['BA IR 2'] = openmc.ZCylinder(r=burn_abs_r2, name='BA IR 2')
surfs['BA IR 3'] = openmc.ZCylinder(r=burn_abs_r3, name='BA IR 3')
surfs['BA IR 4'] = openmc.ZCylinder(r=burn_abs_r4, name='BA IR 4')
surfs['BA IR 5'] = openmc.ZCylinder(r=burn_abs_r5, name='BA IR 5')
surfs['BA IR 6'] = openmc.ZCylinder(r=burn_abs_r6, name='BA IR 6')
surfs['BA IR 7'] = openmc.ZCylinder(r=burn_abs_r7, name='BA IR 7')
surfs['BA IR 8'] = openmc.ZCylinder(r=burn_abs_r8, name='BA IR 8')
surfs['IT IR'] = surfs['BA IR 5']
surfs['IT OR'] = surfs['BA IR 6']
# Rectangular prisms for grid spacers
surfs['rod grid box'] = \
openmc.get_rectangular_prism(rod_grid_side, rod_grid_side)
surfs['rod grid box'] = openmc.rectangular_prism(rod_grid_side, rod_grid_side)
# Rectangular prisms for lattice grid sleeves
surfs['lat grid box inner'] = \
openmc.get_rectangular_prism(17.*pin_pitch, 17.*pin_pitch)
surfs['lat grid box outer'] = \
openmc.get_rectangular_prism(grid_strap_side, grid_strap_side)
surfs['lat grid box inner'] = openmc.rectangular_prism(17.*pin_pitch, 17.*pin_pitch)
surfs['lat grid box outer'] = openmc.rectangular_prism(grid_strap_side, grid_strap_side)
surfs['bot support plate'] = openmc.ZPlane(
z0=bottom_support_plate, name='bot support plate')
surfs['top support plate'] = openmc.ZPlane(
z0=top_support_plate, name='top support plate')
surfs['bot support plate'] = openmc.ZPlane(z0=bottom_support_plate, name='bot support plate')
surfs['top support plate'] = openmc.ZPlane(z0=top_support_plate, name='top support plate')
surfs['bottom FR'] = openmc.ZPlane(z0=bottom_fuel_rod, name='bottom FR')
surfs['top lower nozzle'] = copy.deepcopy(surfs['bottom FR'])
surfs['bot lower nozzle'] = copy.deepcopy(surfs['top support plate'])
surfs['top lower nozzle'] = surfs['bottom FR']
surfs['bot lower nozzle'] = surfs['top support plate']
# axial surfaces
surfs['bot active core'] = openmc.ZPlane(
z0=bottom_fuel_stack, name='bot active core')
surfs['top active core'] = openmc.ZPlane(
z0=top_active_core, name='top active core')
surfs['bot active core'] = openmc.ZPlane(z0=bottom_fuel_stack, name='bot active core')
surfs['top active core'] = openmc.ZPlane(z0=top_active_core, name='top active core')
surfs['top lower thimble'] = copy.deepcopy(surfs['bot active core'])
surfs['BA bot'] = openmc.ZPlane(
z0=bot_burn_abs, name='bottom of BA')
surfs['top lower thimble'] = surfs['bot active core']
surfs['BA bot'] = openmc.ZPlane(z0=bot_burn_abs, name='bottom of BA')
for i, (bottom, top) in enumerate(zip(grid_bottom, grid_top)):
# Create plane for bottom of spacer grid
@ -206,17 +180,12 @@ for i, (bottom, top) in enumerate(zip(grid_bottom, grid_top)):
name = 'top of grid {}'.format(i + 1)
surfs[key] = openmc.ZPlane(z0=top, name=name)
surfs['dashpot top'] = openmc.ZPlane(
z0=step0H, name='top dashpot')
surfs['dashpot top'] = openmc.ZPlane(z0=step0H, name='top dashpot')
surfs['top pin plenum'] = openmc.ZPlane(
z0=top_plenum, name='top pin plenum')
surfs['top FR'] = openmc.ZPlane(
z0=top_fuel_rod, name='top FR')
surfs['bot upper nozzle'] = openmc.ZPlane(
z0=bottom_upper_nozzle, name='bottom upper nozzle')
surfs['top upper nozzle'] = openmc.ZPlane(
z0=top_upper_nozzle, name='top upper nozzle')
surfs['top pin plenum'] = openmc.ZPlane(z0=top_plenum, name='top pin plenum')
surfs['top FR'] = openmc.ZPlane(z0=top_fuel_rod, name='top FR')
surfs['bot upper nozzle'] = openmc.ZPlane(z0=bottom_upper_nozzle, name='bottom upper nozzle')
surfs['top upper nozzle'] = openmc.ZPlane(z0=top_upper_nozzle, name='top upper nozzle')
# Control rod bank surfaces for ARO configuration
for bank in ['A','B','C','D','E',]:
@ -225,50 +194,37 @@ for bank in ['A','B','C','D','E',]:
surfs['bankS{} bot'.format(bank)] = openmc.ZPlane(
z0=step248H, name='CR bankS{} bottom'.format(bank))
surfs['bankA top'] = openmc.ZPlane(
z0=bank_top, name='CR bank A top')
surfs['bankA bot'] = openmc.ZPlane(
z0=bank_bot, name='CR bank A bottom')
surfs['bankB top'] = openmc.ZPlane(
z0=bank_top, name='CR bank B top')
surfs['bankB bot'] = openmc.ZPlane(
z0=bank_bot, name='CR bank B bottom')
surfs['bankC top'] = openmc.ZPlane(
z0=bank_top, name='CR bank C top')
surfs['bankC bot'] = openmc.ZPlane(
z0=bank_bot, name='CR bank C bottom')
surfs['bankD top'] = openmc.ZPlane(
z0=bank_top, name='CR bank D top')
surfs['bankD bot'] = openmc.ZPlane(
z0=bank_bot, name='CR bank D bottom')
surfs['bankA top'] = openmc.ZPlane(z0=bank_top, name='CR bank A top')
surfs['bankA bot'] = openmc.ZPlane(z0=bank_bot, name='CR bank A bottom')
surfs['bankB top'] = openmc.ZPlane(z0=bank_top, name='CR bank B top')
surfs['bankB bot'] = openmc.ZPlane(z0=bank_bot, name='CR bank B bottom')
surfs['bankC top'] = openmc.ZPlane(z0=bank_top, name='CR bank C top')
surfs['bankC bot'] = openmc.ZPlane(z0=bank_bot, name='CR bank C bottom')
surfs['bankD top'] = openmc.ZPlane(z0=bank_top, name='CR bank D top')
surfs['bankD bot'] = openmc.ZPlane(z0=bank_bot, name='CR bank D bottom')
# outer radial surfaces
surfs['core barrel IR'] = openmc.ZCylinder(
R=core_barrel_IR, name='core barrel IR')
surfs['core barrel OR'] = openmc.ZCylinder(
R=core_barrel_OR, name='core barrel OR')
surfs['neutron shield OR'] = openmc.ZCylinder(
R=neutron_shield_OR, name='neutron shield OR')
surfs['core barrel IR'] = openmc.ZCylinder(r=core_barrel_IR, name='core barrel IR')
surfs['core barrel OR'] = openmc.ZCylinder(r=core_barrel_OR, name='core barrel OR')
surfs['neutron shield OR'] = openmc.ZCylinder(r=neutron_shield_OR, name='neutron shield OR')
# neutron shield planes
surfs['neutron shield NWbot SEtop'] = openmc.Plane(
A=1., B=neutron_shield_NWbot_SEtop, C=0., D=0.,
a=1., b=neutron_shield_NWbot_SEtop, c=0., d=0.,
name='neutron shield NWbot SEtop')
surfs['neutron shield NWtop SEbot'] = openmc.Plane(
A=1., B=neutron_shield_NWtop_SEbot, C=0., D=0.,
a=1., b=neutron_shield_NWtop_SEbot, c=0., d=0.,
name='neutron shield NWtop SEbot')
surfs['neutron shield NEbot SWtop'] = openmc.Plane(
A=1., B=neutron_shield_NEbot_SWtop, C=0., D=0.,
a=1., b=neutron_shield_NEbot_SWtop, c=0., d=0.,
name='neutron shield NEbot SWtop')
surfs['neutron shield NEtop SWbot'] = openmc.Plane(
A=1., B=neutron_shield_NEtop_SWbot, C=0., D=0.,
a=1., b=neutron_shield_NEtop_SWbot, c=0., d=0.,
name='neutron shield NEtop SWbot')
# outer radial surfaces
surfs['RPV IR'] = openmc.ZCylinder(
R=rpv_IR, name='RPV IR')
surfs['RPV OR'] = openmc.ZCylinder(
R=rpv_OR, name='RPV OR', boundary_type='vacuum')
surfs['RPV IR'] = openmc.ZCylinder(r=rpv_IR, name='RPV IR')
surfs['RPV OR'] = openmc.ZCylinder(r=rpv_OR, name='RPV OR', boundary_type='vacuum')
# outer axial surfaces
surfs['upper bound'] = openmc.ZPlane(