Alexander Vasiliev

I. The core management system (CMSYS) of the Swiss Simulation Platform [1] developed by Laboratory for Reactor Physics and Thermal-Hydraulics (LRT) at Paul Scherrer Institute (PSI) represents high-...

This paper presents the work performed in the subgroup 16 of the Working Party for Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency. The main goal was to define two decay heat benchmarks for Spent Nuclear Fuel (one pincell and one assembly), perform calculations and compare and analyze the results in light of existing calorimetric measurements. The selected case is the PWR UO ₂ assembly 0E2, irradiated at the Ringhals-3 reactor and measured at the Clab facility in Sweden. In total, 21 institutes worldwide participated to the exercise, leading to 55 calculated results (named C ). It was found that the measured decay heat values ( E ) can be satisfactorily reproduced with two-dimensional assembly calculations, leading to an average C / E value of 0.99, with an uncertainty (or one standard deviation) of ±0.01.

In this paper, we are estimating biases for the reactivity decrement and burnup of irradiated fuel assemblies, based on differences between measured and calculated local in-core power measurements (flux maps). The simulations are performed with validated CASMO5 and SIMULATE5 cycle follow-up models for three Pressurized Water Reactor (PWR) cores. A total of more than 70 cycles of operation and 1000 in-core flux maps at various core burnup steps are used, including both UO ₂ and MOX fuel assemblies, as well as UO ₂ alone, with relatively high assembly average burnup at the end of life (greater than 60 MWd/kg). The effect of nuclear data uncertainties is also calculated and quantified on both bias types. The main findings are that while the burnup biases are globally constant as a function of the assembly burnup, the bias for the reactivity decrement tends to increase, both in terms of average values and tolerance intervals.

We analyze in-core reaction rate measurements coming from three Pressurized Water Reactors and two Boiling Water Reactors (named E ) and compare them with CASMO5 and SIMULATE-3 (and -5) calculations (named C ). A total of almost 1 million experimental values were considered, covering various cycles, detector systems and locations, as well as UO ₂ and MOX fuel types. Biases and root mean square values (rms) were extracted based on ( E − C )/ C ratios, leading to the finding of outliers, most likely due to detector issues and possible modeling effects for locations with low power. The average rms value is 2.9%, once the outliers are removed, but dissimilar values are obtained for different reactor types. The effect of nuclear data is also quantified for one PWR, leading to an absolute average rms uncertainty of 0.3%. We conclude, based on this study, that an improvement of the local rms values can be achieved by a better knowledge of measurements ( i.e. uncertainties) and a possible update on specific parts of the involved models.