Criticality benchmark of McCARD Monte Carlo code for light-water-reactor fuel in transportation and storage packages

In this paper, McCARD code was verified using various models listed in the NUREG/CR-6361 benchmark guide, which provides specifications for single pin-cells, single assemblies, and the whole core classified depending on the nuclear properties and structural characteristics. McCARD code was verified by comparing its results with those of SCALE code for single pin-cell and single assembly benchmark problems. The difference in the multiplication factor obtained through the two codes did not exceed 90 pcm. The benchmark guide treats a total of 173 whole core experiments. The experiments are categorized as simple lattices, separator plates, reflecting walls, reflecting walls and separator plates, burnable absorber fuel rods, water holes, poison rods, and borated moderator. As a result of numerical simulation using McCARD, the mean value of the multiplication factors is 1.00223 and the standard deviation of the multiplication factors is 285 pcm. The difference between the multiplication factors and the experimental value is in the range of -665 pcm to + 1609 pcm. In addition, statistics of results for experiments categorized by reactor shape, additional structure, burnable poison, etc., are detailed in the main text. Keywords: McCARD, LWR benchmar

Criticality benchmark of Monte Carlo code MCS for light water reactor fuel in transportation and storage packages

This paper presents a verification and validation (V&V) study of the Monte Carlo neutron transport code MCS for the criticality analysis of LWR fuel in transportation and storage casks/packages. A total of 173 critical experiments from the NUREG/CR-6361 report are analyzed with MCS and the ENDF/B-VII.1 cross section library. A preliminary verification is conducted by comparing the reaction rates. Subsequently, a criticality analysis of the 173 benchmarks is performed. The average value of the multiplication factors calculated by MCS for the 173 critical experiments is 1.00243 with a standard deviation of 291 pcm and extremum error values of -1451 pcm and +1661 pcm compared to the experimental results. The bias of the package (MCS + ENDF/B-VII.1 library) for the criticality analysis of LWR fuel transportation/storage casks is, therefore, estimated to be +243 pcm. This paper demonstrates the capability of the MCS code to perform the criticality analysis of LWR fuel transportation and storage packages. (C) 2020 Elsevier Ltd. All rights reserved

Verification of Graphite Isotope Ratio Method Combined With Polynomial Regression for the Estimation of Cumulative Plutonium Production in a Graphite-Moderated Reactor

Graphite Isotope Ratio Method (GIRM) can be used to estimate plutonium production in a graphite-moderated reactor. This study presents verification results for the GIRM combined with a 3-D polynomial regression function to estimate cumulative plutonium production in a graphite-moderated reactor. Using the 3-D Monte-Carlo method, verification was done by comparing the cumulative plutonium production with the GIRM. The GIRM can estimate plutonium production for specific sampling points using a function that is based on an isotope ratio of impurity elements. In this study, the 10B/11B isotope ratio was chosen and calculated for sampling points. Then, 3-D polynomial regression was used to derive a function that represents a whole core cumulative plutonium production map. To verify the accuracy of the GIRM with polynomial regression, the reference value of plutonium production was calculated using a Monte-Carlo code, MCS, up to 4250 days of depletion. Moreover, the amount of plutonium produced in certain axial layers and fuel pins at 1250, 2250, and 3250 days of depletion was obtained and used for additional verification. As a result, the difference in the total cumulative plutonium production based on the MCS and GIRM results was found below 3.1% with regard to the root mean square (RMS) error