Sensitivity and uncertainty analysis on the criticality by an ERRORJ/SUSD3D with JENDL-3.3 covariance data

The covariance data processing and the nuclear data sensitivity and uncertainty analysis of the keff have been carried out for some 1-D benchmark problems by using the ERRORJ/SUSD3D code system. The uncertainties due to the U-233, U-235, U-238, Pu-239, Pu-240, and/or Pu-241 covariance data of JENDL-3.3 have been estimated with the P3-S16 and P3-S4 approximations. The uncertainties amount to 0.53% ~ 1.86% depending on the major actinides of the benchmark cores

A SIMPLE METHOD TO CALCULATE THE DISPLACEMENT DAMAGE CROSS SECTION OF SILICON CARBIDE

We developed a simple method to prepare the displacement damage cross section of SiC using NJOY and SRIM/TRIM. The number of displacements per atom (DPA) dependent on primary knock-on atom (PKA) energy was computed using SRIM/TRIM and it is directly used by NJOY/HEATR to compute the neutron energy dependent DPA cross sections which are required to estimate the accumulated DPA of nuclear material. SiC DPA cross section is published as a table in DeCART 47 energy group structure. Proposed methodology can be easily extended to other materials

Fission yields data generation and benchmarks of decay heat estimation of a nuclear fuel

Fission yields data with the ENDF-6 format of 235U, 239Pu, and several actinides dependent on incident neutron energies have been generated using the GEF code. In addition, fission yields data libraries of ORIGEN-S, -ARP modules in the SCALE code, have been generated with the new data. The decay heats by ORIGEN-S using the new fission yields data have been calculated and compared with the measured data for validation in this study. The fission yields data ORIGEN-S libraries based on ENDF/B-VII.1, JEFF-3.1.1, and JENDL/FPY-2011 have also been generated, and decay heats were calculated using the ORIGEN-S libraries for analyses and comparisons

Fission yields data generation and benchmarks of decay heat estimation of a nuclear fuel

Fission yields data with the ENDF-6 format of 235U, 239Pu, and several actinides dependent on incident neutron energies have been generated using the GEF code. In addition, fission yields data libraries of ORIGEN-S, -ARP modules in the SCALE code, have been generated with the new data. The decay heats by ORIGEN-S using the new fission yields data have been calculated and compared with the measured data for validation in this study. The fission yields data ORIGEN-S libraries based on ENDF/B-VII.1, JEFF-3.1.1, and JENDL/FPY-2011 have also been generated, and decay heats were calculated using the ORIGEN-S libraries for analyses and comparisons

Calculation of fission product yields for uranium isotopes by using a semi-empirical model

A semi-empirical model for calculating the fission product yields (FPY) of neutron induced fissions of uranium isotopes is developed, where the FPY are assumed to be proportional to the level density of a microcanonical ensemble of a compound nucleus at the fission barrier. The fission height that determines the level density is modeled as a sum of two parts; a symmetric part and an asymmetric part. The origin of the symmetric part can be attributed to the liquid drop model, and that of the asymmetric part to the shell effect in the fission products. Our model has essentially just seven adjustable parameters. They are fitted to the ENDF/B-VII.1 fission yield data of various uranium isotopes for the mass number ranging from 232 to 238 induced by thermal and fast (500 keV) neutrons. Five of the resulting parameters are nearly independent of the mass number of the uranium isotopes. Two parameters which change with the mass number of the uranium isotopes can be expressed as a linear function of the mass number. The FPY calculated from our model are found to be in a good agreement with both the ENDF and experimental data