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.
Highlights
The present analysis is the continuation of the multi-year effort started at the Paul Scherrer Institute (PSI) for a better understanding of the neutronics simulation capabilities of nuclear fuel
The decay heat is an important quantity as it plays a keyrole in the design of canisters and geological repositories for long-term spent fuel storage, and nuclide inventory is directly linked to the risk of criticality of such canisters
Very similar works were recently performed for the GU1 and GU3 samples [5,6]. These two samples are made of UO2 fuel, whereas the BM1 and BM3 samples are made of Mixed OXyde; the analysis of these four samples will eventually lead to more general conclusions regarding decay heat and nuclide concentration calculations based on the CASMO5, SIMULATE and SNF codes [7,8,9]
Summary
The present analysis is the continuation of the multi-year effort started at the Paul Scherrer Institute (PSI) for a better understanding of the neutronics simulation capabilities of nuclear fuel Such attempt is related to the current studies performed in the European Union project called EURAD, and more specially its work package 8 [1], and within the current Coordinated Research Project (or CRP) on Spent Fuel Characterization from the IAEA [2]. Very similar works were recently performed for the GU1 and GU3 samples (same simulation process and same uncertainty method) [5,6] These two samples are made of UO2 fuel, whereas the BM1 and BM3 samples are made of Mixed OXyde (or MOX); the analysis of these four samples will eventually lead to more general conclusions regarding decay heat and nuclide concentration calculations based on the CASMO5, SIMULATE and SNF codes [7,8,9]. A number of numerical results are not directly included in the sections of this paper, but can be found in Appendix
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