Problem with 'diff_burnable_mats' in the depletion of fuel assembly

User Support
1

Summary

  • OpenMC - 0.13v in conda environment
  • ‘diff_burnable_mats = True’ used for considering repeated materials
  • but it seems depletable materials are not reproduced in materials.xml
  • wonder if missing options in the input for materials

Hi there,

I am glad to post my first topic and say hello to you :slight_smile: Since I have been looking into OpenMC code only for a few days, I am afraid that my concern could be so elementary and simple. I tried to figure out a problem and find a solution through a forum and documents, but I couldn’t. So your comments would be really helpful and appreciated.

I am trying to simulate a simple problem, the depletion of a single fuel assembly, to catch up functions and get used to the code. I suffered the problem when using diff_burnable_mats = True in order to properly consider repeated materials. The problem is that depletable materials were not reproduced.

plot_1
plot_11200×1200 25.1 KB

Fuel assembly of PWR

I only defined five materials such as UO2, Gad, Clad, Tube and moderator. UO2 and Gad are depletable. In geometry.xml, materials numbering over 6 would be automatically generated materials.

geometry.xml

  2 <geometry>
  3   <cell id="1" material="6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 4    2 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85     86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 101 102 103 104 105 106 107 108 109 110 111 112 113 114 115 116 117 118 119 120 121     122 123 124 125 126 127 128 129 130 131 132 133 134 135 136 137 138 139 140 141 142 143 144 145 146 147 148 149 150 151 152 153 1    54 155 156 157 158 159 160 161 162 163 164 165 166 167 168 169 170 171 172 173 174 175 176 177 178 179 180 181 182 183 184 185 186     187 188 189 190 191 192 193 194 195 196 197 198 199 200 201 202 203 204 205 206 207 208 209 210 211 212 213 214 215 216 217 218 2    19 220 221 222 223 224 225 226 227 228 229 230 231 232 233 234 235 236 237 238 239 240 241 242 243 244 245 246 247 248 249 250 251     252 253" region="-11" universe="1" />
  4   <cell id="2" material="3" region="11 -12" universe="1" />
  5   <cell id="3" material="5" region="12" universe="1" />
  6   <cell id="4" material="254 255 256 257 258 259 260 261 262 263 264 265 266 267 268 269" 
 12   <cell id="10" material="350 351 352 353 354 355 356 357 358 359 360 361 362 363 364 365" region="20 -11" universe="2" />
 13   <cell id="11" material="3" region="11 -12" universe="2" />
 16   <cell id="14" material="4" region="13 -14" universe="3" />

python

dp = openmc.model.Model(geometry, materials_file, settings_file)
op = openmc.deplete.Operator(dp, chain_file, diff_burnable_mats=True)

celi = openmc.deplete.CELIIntegrator(op, time_steps, power, timestep_units='MWd/kg')
celi.integrate()

.
.
But in materials.xml, only five materials that I defined were specified. When I run, I got the following error message, of course.

                 | The OpenMC Monte Carlo Code
       Copyright | 2011-2022 MIT, UChicago Argonne LLC, and contributors
         License | https://docs.openmc.org/en/latest/license.html
         Version | 0.13.0
        Git SHA1 | d9478c1304a37590453eabe3a99e97abb7f92013
       Date/Time | 2022-06-28 12:01:28
   MPI Processes | 1
  OpenMP Threads | 20

 Reading settings XML file...
 Reading cross sections XML file...
 Reading materials XML file...
 Reading geometry XML file...
 ERROR: Could not find material 6 specified on cell 1

I am wondering if I miss some options to materials for the depletion calculation. You can see details in the attachments.

materials.xml (2.7 KB)
geometry.xml (4.5 KB)
run_depletion.py (8.7 KB)

2

Continued…

After looking into further, I presume that the reason my depletable materials were not cloned by diff_burnable_mats was because I did not define my cell type to be ‘distribmat’.

So here are questions.

(1) Should I define the cell type to be ‘distribmat’ for openmc.Cell() variable when using diff_burnable_mats?

(2) In addition, could num_instances in openmc.Cell() be related with the problem?

(3) Why can’t I define the cell type with the variable input?
When I tried to specify the cell type, I got an error.

pin = openmc.Cell()
pin.fill_type = 'distribmat'

pin.fill_type = ‘distribmat’
AttributeError: can’t set attribute ‘fill_type’

I would appreciate your help. Thanks.

3

@Inhyung There was actually a bug with diff_burnable_mats that may be affecting your particular situation. That bug has been fixed since the last release of OpenMC and will be included in our next release (which will be out very soon).

No, when you use diff_burnable_mats, OpenMC will automatically take care of setting up distributed materials for you.

The Cell.num_instances attribute is set if you call Geometry.determine_paths(). This is done automatically by the diff_burnable_mats option.

The fill_type attribute is read-only and determined by whatever you fill the material with. If you were to set, e.g.,

pin.fill = [mat1, mat2, ...]

You’ll see afterward that pin.fill_type will be “distribmat”.

1 Like
4

@paulromano Thank you so much for your kind reply. Now I got all the questions and the current situation. It really helped me a lot. I will look forward to the new version. Thanks again.