# Depletion failure **URL:** \ **Category:** User Support\ **Created:** [January 12, 2023, 8:59pm UTC](https://openmc.discourse.group/t/depletion-failure/2614 "2023-01-12T20:59:41Z")\ **Posts on this page:** 2\ **Page:** 1
**Author:** ![Brian729](https://avatars.discourse-cdn.com/v4/letter/b/a8b319/32.png) [@Brian729](https://openmc.discourse.group/u/Brian729)\ **Post date:** [January 12, 2023, 8:59pm UTC](https://openmc.discourse.group/t/depletion-failure/2614/1 "2023-01-12T20:59:41Z")
I am running a U233/Th232 molten salt reactor simulation. I am using the following depletion code: Power is 3 Gigawatts time\_steps\_day=[24_60_60]\*10 model = openmc.Model(geometry=geometry,materials=materials, settings=settings) operator = openmc.deplete.CoupledOperator(model,chainfile) integrator = openmc.deplete.PredictorIntegrator(operator, time\_steps\_day, power) integrator.integrate() results = openmc.deplete.ResultsList.from\_hdf5(“./depletion\_results.h5”) I get the following warning after 6 depletion iterations: Creating state point openmc\_simulation\_n6.h5… /home/brian/.local/lib/python3.10/site-packages/openmc/deplete/helpers.py:427: RuntimeWarning: overflow encountered in scalar divide return source\_rate / energy /home/brian/.local/lib/python3.10/site-packages/openmc/deplete/openmc\_operator.py:548: RuntimeWarning: invalid value encountered in multiply rates \*= self.\_normalization\_helper.factor(source\_rate) /home/brian/.local/lib/python3.10/site-packages/scipy/sparse/linalg/\_dsolve/linsolve.py:276: MatrixRankWarning: Matrix is exactly singular warn(“Matrix is exactly singular”, MatrixRankWarning) /home/brian/.local/lib/python3.10/site-packages/scipy/sparse/linalg/\_dsolve/linsolve.py:276: MatrixRankWarning: Matrix is exactly singular warn(“Matrix is exactly singular”, MatrixRankWarning) And the following failure report: AllocationError Traceback (most recent call last) Cell In[16], line 5 3 operator = openmc.deplete.CoupledOperator(model,chainfile) 4 integrator = openmc.deplete.PredictorIntegrator(operator, time\_steps\_day, power) ----\> 5 integrator.integrate() 6 results = openmc.deplete.ResultsList.from\_hdf5(“./depletion\_results.h5”) File ~/.local/lib/python3.10/site-packages/openmc/deplete/abc.py:806, in Integrator.integrate(self, final\_step, output) 804 # Solve transport equation (or obtain result from restart) 805 if i \> 0 or self.operator.prev\_res is None: → 806 conc, res = self.\_get\_bos\_data\_from\_operator(i, source\_rate, conc) 807 else: 808 conc, res = self.\_get\_bos\_data\_from\_restart(i, source\_rate, conc) File ~/.local/lib/python3.10/site-packages/openmc/deplete/abc.py:759, in Integrator.\_get\_bos\_data\_from\_operator(self, step\_index, source\_rate, bos\_conc) 756 “”“Get beginning of step concentrations, reaction rates from Operator 757 “”” 758 x = deepcopy(bos\_conc) → 759 res = self.operator(x, source\_rate) 760 self.operator.write\_bos\_data(step\_index + self.\_i\_res) 761 return x, res File ~/.local/lib/python3.10/site-packages/openmc/deplete/coupled\_operator.py:465, in CoupledOperator. **call** (self, vec, source\_rate) 462 openmc.lib.run() 464 # Extract results → 465 rates = self.\_calculate\_reaction\_rates(source\_rate) 467 # Get k and uncertainty 468 keff = ufloat(\*openmc.lib.keff()) File ~/.local/lib/python3.10/site-packages/openmc/deplete/openmc\_operator.py:533, in OpenMCOperator.\_calculate\_reaction\_rates(self, source\_rate) 530 for nuc, i\_nuc\_results in zip(rxn\_nuclides, nuc\_ind): 531 number[i\_nuc\_results] = self.number[mat, nuc] → 533 tally\_rates = self.\_rate\_helper.get\_material\_rates( 534 mat\_index, nuc\_ind, react\_ind) 536 # Compute fission yields for this material 537 fission\_yields.append(self.\_yield\_helper.weighted\_yields(i)) File ~/.local/lib/python3.10/site-packages/openmc/deplete/helpers.py:217, in DirectReactionRateHelper.get\_material\_rates(self, mat\_id, nuc\_index, react\_index) 198 “”“Return an array of reaction rates for a material 199 200 Parameters (…) 214 reaction rates in this material 215 “”” 216 self.\_results\_cache.fill(0.0) → 217 full\_tally\_res = self.rate\_tally\_means[mat\_id] 218 for i\_tally, (i\_nuc, i\_react) in enumerate( 219 product(nuc\_index, react\_index)): 220 self.\_results\_cache[i\_nuc, i\_react] = full\_tally\_res[i\_tally] File ~/.local/lib/python3.10/site-packages/openmc/deplete/helpers.py:186, in DirectReactionRateHelper.rate\_tally\_means(self) 184 # If the mean cache is empty, fill it once with this transport cycle’s results 185 if self.\_rate\_tally\_means\_cache is None: → 186 self.\_rate\_tally\_means\_cache = self.\_rate\_tally.mean 187 return self.\_rate\_tally\_means\_cache File ~/.local/lib/python3.10/site-packages/openmc/lib/tally.py:276, in Tally.mean(self) 273 @property 274 def mean(self): 275 n = self.num\_realizations → 276 sum\_ = self.results[:, :, 1] 277 if n \> 0: 278 return sum\_ / n File ~/.local/lib/python3.10/site-packages/openmc/lib/tally.py:306, in Tally.results(self) 304 data = POINTER(c\_double)() 305 shape = (c\_size\_t\*3)() → 306 \_dll.openmc\_tally\_results(self.\_index, data, shape) 307 return as\_array(data, tuple(shape)) File ~/.local/lib/python3.10/site-packages/openmc/lib/error.py:21, in \_error\_handler(err, func, args) 19 # Raise exception type corresponding to error code 20 if err == errcode(‘OPENMC\_E\_ALLOCATE’): —\> 21 raise exc.AllocationError(msg) 22 elif err == errcode(‘OPENMC\_E\_OUT\_OF\_BOUNDS’): 23 raise exc.OutOfBoundsError(msg) AllocationError: Tally results have not been allocated yet. I have tried varying source particle settings, varying the duration of the time steps, neither of which made any difference except when I changed the time step duration to a week, it failed one second iteration. Keff is well over 1.0 (1.03) when the failure occurs. Any suggestions? ---
**Author:** ![KUNG](https://yyz2.discourse-cdn.com/free1/user_avatar/openmc.discourse.group/kung/32/3725_2.png) [@KUNG](https://openmc.discourse.group/u/KUNG)\ **Post date:** [October 28, 2024, 5:01pm UTC](https://openmc.discourse.group/t/depletion-failure/2614/2 "2024-10-28T17:01:25Z")
_Have you figured out what caused your “AllocationError: Tally results have not been allocated yet.”?_ This error appeared for me when I tried to run a pincell depletion example with an erroneous chain file (jeff33 chain file I generated myself with missing fission reactions and yields)… Without fission reactions/yields in the chain file I guess it is not surprising that it will fail after the the first transport step. With the corrected chain file the depletion example runs. **Error in pincell depletion example with erroneous jeff33 chain:** Traceback (most recent call last): File “~/openmc/calculations/AP1/pincell\_depletion\_07/pincell\_depletion.py”, line 360, in integrator.integrate() File “~/openmc/venv-openmc/lib/python3.10/site-packages/openmc/deplete/abc.py”, line 801, in integrate n, res = self.\_get\_bos\_data\_from\_operator(i, source\_rate, n) File “~/openmc/venv-openmc/lib/python3.10/site-packages/openmc/deplete/abc.py”, line 745, in \_get\_bos\_data\_from\_operator res = self.operator(x, source\_rate) File “~/openmc/venv-openmc/lib/python3.10/site-packages/openmc/deplete/coupled\_operator.py”, line 456, in **call** rates = self.\_calculate\_reaction\_rates(source\_rate) File “~/openmc/venv-openmc/lib/python3.10/site-packages/openmc/deplete/openmc\_operator.py”, line 527, in \_calculate\_reaction\_rates tally\_rates = self.\_rate\_helper.get\_material\_rates( File “~/openmc/venv-openmc/lib/python3.10/site-packages/openmc/deplete/helpers.py”, line 219, in get\_material\_rates full\_tally\_res = self.rate\_tally\_means[mat\_index] File “~/openmc/venv-openmc/lib/python3.10/site-packages/openmc/deplete/helpers.py”, line 188, in rate\_tally\_means self.\_rate\_tally\_means\_cache = self._rate\_tally.mean File “~/openmc/venv-openmc/lib/python3.10/site-packages/openmc/lib/tally.py”, line 310, in mean sum_ = self.results[:, :, 1] File “~/openmc/venv-openmc/lib/python3.10/site-packages/openmc/lib/tally.py”, line 340, in results \_dll.openmc\_tally\_results(self.\_index, data, shape) File “~/openmc/venv-openmc/lib/python3.10/site-packages/openmc/lib/error.py”, line 21, in \_error\_handler raise exc.AllocationError(msg) openmc.exceptions.AllocationError: Tally results have not been allocated yet. **The script I used to generate the jeff33 chain file is an adapted one of this one here:** [data/depletion/generate\_tendl\_chain.py at master · openmc-dev/data · GitHub](https://github.com/openmc-dev/data/blob/master/depletion/generate_tendl_chain.py) **Update:** My jeff33 chain file was erroneous (fission product yield and fission reactions were missing since I pointed incorrectly to the fpy\_files). **The warnings/messages from _openmc.deplete.Chain.from\_endf_ for jeff33 were:** […] [Various warnings from decay.py about continuous spectra with log-linear interpolations e.g. for Th230, Th232, … Es 253] […] Decay mode [‘beta-’] of parent In123\_m1 has a zero branching ratio. […] The following fissionable nuclides have no fission product yields: Ra223, replaced with U235 Ra226, replaced with U235 Ac225, replaced with U235 […] Fm255, replaced with U235