Validation of BWR spent nuclear fuel isotopic predictions with applications to burnup credit

Multiphysics Modeling and Validation of Spent Fuel Isotopics Using Coupled Neutronics/Thermal-Hydraulics Simulations

“…e average relative difference for all isotopes can be less than 15% for some samples but can also be higher than that for other samples. It is worth mentioning that the current validation results are similar to previous efforts such as [4,5,8]. 10 0 10 -1 10 -2 10 -3 10 -4 10 -5 Experimental (g/gHmi) Finally, since lack of documentation was a major challenge for this work, future efforts will focus on incorporating machine learning and data science to aid in identifying the missing input parameters such that they can yield the desired output.…”

“…For BWRs, even with reported experimental data, major assumptions were needed to be made about the behavior of axial coolant density distributions in order to create verification models in [4]. However, such assumptions can introduce uncertainties in the calculations because void fraction feedback is not captured [5,6]. e current study seeks to avoid this difficulty by coupling the thermal-hydraulics code TRACE to the T5-DEPL depletion sequence in SCALE for validation of spent fuel composition, thereby avoiding the need for detailed history of void distribution data.…”

“…Many methods of uncertainty quantification based on Monte Carlo parametric sampling [3,9,10] exist; data on C/E can be helpful in approximating uncertainty bounds of the isotopic concentrations in spent fuel analysis, similar to what has been done in [11]. An uncertainty analysis of a spent nuclear fuel storage system was performed in [5]. POLARIS was used to calculate the composition of commercial BWR spent fuel for 77 samples.…”

“…is study [5] presented and compared BWR spent fuel isotopics for both actinides and fission products. en, the study compared k eff of a spent fuel cask based on codepredicted isotopic compositions with a cask based on experimentally reported isotopic compositions.…”

Analysis with the ORIGEN Module of PNAR Spent Fuel Measurements for Nuclear Safeguards Applications

“…The total Pu is predicted with an average bias of less than 1% and standard deviation of 5%. The 235 U results exhibit larger deviations in this study compared with spent fuel measurements [11], where 235 U was overpredicted by ~4% on average compared with measurements; however, the present calculations do not apply any detailed design or operating history information of the assemblies.…”