Analysis for the ARIANE BM1 and BM3 samples: nuclide inventory and decay heat

Abstract

The Mixed Oxide samples (MOX) ARIANE Post Irradiation Examination samples BM1 and BM3 have been analyzed in this work, based on various two- and three-dimensional models. Calculated and measured nuclide inventories are compared based on CASMO5, SIMULATE and SNF simulations, and calculated values for the decay heat of the assembly containing the samples are also provided. For uncertainty propagation, the covariance information from three different nuclear data libraries are used. Uncertainties from manufacturing tolerances and operating conditions are also considered. The results from these two samples are compared with the ones from two UO2 samples, namely GU1 and GU3, also from the ARIANE program, applying the same calculation scheme and uncertainty assumptions. It is shown that a two-dimensional assembly model provides better agreement with the measurements than a two-dimensional single pin model, and that the full core three-dimensional model provides similar results compared to the assembly model, although no 148Nd normalization is applied for the full core model. For the MOX assembly decay heat, as expected, heavy actinides have a higher contribution compared to the cases with the UO2 samples; additionally, decay heat uncertainties are moderately smaller in the case of the MOX assembly.

Analysis of ENRESA BWR samples: nuclide inventory and decay heat

“…Values are presented in Table A. 5. One can see that naturally uncertainties are increased, notably at short cooling time (up to 17% at 1 year cooling time).…”

“…It should be pointed out that, in realistic cases without measurements, the normalization to 148 Nd will not be available and code performances cannot be expected to be as good as when the 148 Nd concentrations are known. The impact of not normalizing the sample burnup was also presented in reference [5] for the BM1 sample and full core simulations.…”

“…This way, uncertainties are obtained at the same time for all quantities calculated by CASMO5: nuclide concentrations as well as decay heat. Apart for the void fraction, all input parameters and their uncertainties were varied in a similar way (same method, same initial variations) for the PWR samples called GU1, GU3 and BM1, presented in references [3][4][5], allowing a consistent comparison of uncertainties.…”

“…The present work is also part of the validation of specific calculation scheme, representing the modelization of a specific irradiation environment (for instance a two-or three-dimensional assembly), the irradiation history (cycle parameters, cooling time), boundary conditions and finally the simulation tool (including neutronic transport, depletion and decay). Recently, a consistent method was applied to the Pressurized Water Rector (PWR) UO 2 samples GU1 and GU3 and to the PWR MOX samples BM1 and BM3 [3][4][5] (similar hypothesis with respect to the simulation codes and methods, nuclear data libraries, uncertainties), allowing to compare the prediction capabilities, as well as calculated uncertainties.…”

Spent nuclear fuel management, characterisation, and dissolution behaviour: progress and achievement from SFC and DisCo

“…For decay heat, the calorimetric measurements from CLAB (Swedish Central Intermediate Storage Facility), GE-Morris, and HEDL (Hanford Engineering Development Laboratory) were considered, leading to more than 250 cases. To support the recommendations of this subtask, some studies were already published, see references [3][4][5][6][7][8][9][10][11][12][13][14][15]. A global review of the The aim is to produce NDA systems that can be used to provide experimental data to validate the performance of the codes (including nuclear data) and to verify the declared design and operational history of the assemblies.…”