Akihiro Sasahara

Development of Partitioning and Transmutation Technology for Long-Lived Nuclides

High-level radioactive wastes from the Purex process contain long-lived nuclides, which mainly consist of transuranic elements (TRUs). A partitioning and transmutation method that consists of pyrom...

Neutron and Gamma Ray Source Evaluation of LWR High Burn-up UO2 and MOX Spent Fuels

The axial neutron emission and gamma ray source distribution were measured for LWR high burn-up UO2 and MOX spent fuel rods. The gamma rays of 134Cs, 137Cs and 106Ru were measured on the fuel rods, and consequently compared with the results of the ORIGEN2/82 calculation, in which both the original library and ORLIBJ32 based on the JENDL-3.2 library were used. The effect of pellet radial gamma distribution on self-shielding factor was considered, and cesium radial migration was also discussed using the ratio of gamma ray intensity of 134Cs to 137Cs. The axial gamma ray measurement and calculation of 134Cs and 106Ru agreed within about 20% and 137Cs agreed within 13% after correction with pellet self-shielding factors. Two types of boundary curves associated with cesium migration were obtained by simple analysis using 134Cs/137Cs gamma intensity ratio. The axial neutron emission calculated by ORIGEN2/82 with ORLIBJ32 was generally smaller than the measured ones. The main neutron source in spent fuel is the spontaneous fission of 244Cm. Small buildup of 244Cm caused the underestimation of neutron emission. The main flow up to 244Cm is 241 Pu, 242Pu, 243Pu, 243 Am and 2440Am. The cross sections of 243 Am on the main flow and 244Cm have strong sensitivity on buildup of 244Cm. The re-evaluation of the cross sections of these nuclides will improve the prediction for 244Cm buildup.

Core performance of fast reactors for actinide recycling using metal, nitride, and oxide fuels

Core performance analyses are conducted for fast reactors that accept and recycle the plutonium and minor actinides (MAs) recovered from light water reactor (LWR) spent fuel, together with the plutonium and MAs from the fast reactors` own production. Metal, nitride, and oxide are the fuel materials used to compare the neutronic and safety parameters and to discuss acceptable minor actinide content. Based on the material balance of the analyzed cores, an LWR-fast reactor fuel cycle model is used to calculate the mass flow of the plutonium and MAs and to estimate their total amount in the waste stream.

Chemical Isotopic Analysis of Fission Products in PWR-MOX Spent Fuels and Computational Evaluation Using JENDL, ENDF/B, JEF, and JEFF

A chemical isotopic analysis of high-burn-up MOX spent fuels with a burn-up of 45 MWd/kgHM was carried out to accumulate nuclide composition data. Furthermore, computational analysis was performed using the integrated burn-up calculation code system SWAT. The differences between the amounts of fission products obtained by the chemical isotopic analysis and SWAT calculation using JENDL-3.2, JENDL-3.3, ENDF/B-VI.5, ENDF/B-VI.8, JEF-2.2, and JEFF-3.0 were evaluated as the ratios of the calculated values to the experimental ones (C/E ratios). The fission products such as 88Sr, 90Sr, 106Ru, 133Cs, 134Cs, and 135Cs, which are gamma and decay heat sources, neutron absorption nuclides, or burn-up indicators in spent fuels, were further investigated to improve their C/E ratios using the simplified burn-up chains of fission products using JENDL-3.3; consequently, the correction values for the fission yields or capture cross sections of the fission products were estimated using sensitivity coefficients. The C/E ratios for 154Eu, 155Eu, 154Gd, 155Gd, and 156Gd markedly differed among libraries. The reason for this difference was also discussed using the sensitivity coefficient and capture cross section of each fission product, which is in their main sensitive production paths.

The Long-Term Performance of Concrete in Nuclear Applications

The Idaho National Laboratory investigated long-term concrete performance for nuclear applications. Scientists searched the open literature for information on the effects of heat and radiation. They also examined the concrete shield of the Pacific Sierra Nuclear Ventilated Storage Cask (VSC-17) for concrete deterioration and loss of shielding. The literature search revealed that low doses, <109 n/cm2 or <1011 Rad gamma, of radiation over periods less than 50 years do not appear to significantly affect concrete. Exposure over 100 years was not studied. The effects of higher doses of radiation are not as clear. Generally, the threshold of degradation is 95°C, and degradation increases with increasing temperature and time. In 15 years of VSC-17 use, there is no apparent effect from environment, radiation, or temperature that has adversely affected the shielding or structural functions, although there has been some minor cracking of the concrete.© 2005 ASME

Isotopic Analysis of Actinides and Fission Products in LWR High-Burnup UO2 Spent Fuels and its Comparison with Nuclide Composition Calculated Using JENDL, ENDF/B, JEF and JEFF

A chemical isotopic analysis of the actinides and fission products of a high-burnup PWR-UO2 fuel with an average burnup of 60.2 MWd/kgHM was carried out to accumulate extensive nuclide composition data. Furthermore, computational analysis was performed using the integrated burnup calculation code SWAT. The differences between the amounts obtained by the chemical isotopic analysis and SWAT calculation using JENDL-3.2, JENDL-3.3, ENDF/B-VI.5, ENDF/B-VI.8, JEF-2.2 and JEFF-3.0 were evaluated as the ratios of the calculated values to the experimental values (C/E ratios). For actinides, the calculated 244Cm amount, which is an important nuclide as a major neutron source in spent fuels, was underestimated. The main sensitive path for 244Cm was therefore investigated by a simple depletion calculation for actinides and the cause of the underestimation of the calculated 244Cm amount is discussed. The fission products 88Sr, 90Sr, 89Y, 106Ru, 133Cs, 135Cs and 144Nd, for which the C/E ratios were larger than 1.05 or smaller than 0.95, were investigated to improve their C/E ratios by a simple depletion calculation with simplified burnup chains and sensitivity coefficients, and the correction values for the fission yields or capture cross sections were estimated. The effect of power history on nuclide composition was also investigated. Additionally, the fission products for which the C/E ratios strongly depend on the type of library are discussed using sensitivity coefficients.

Neutron/Gamma Ray Source Measurement and Analyses of High Burnup UO2/MOX Fuel Rods

Axial directional neutron and gamma ray source measurement of fuel rods and radial directional gamma ray source measurement of fuel pellets were carried out for LWR high burn-up UO2 and MOX spent fuels. C/E values with ORIGEN-2 code are discussed for axial neutron and gamma ray source distributions of fuel rods. The radial gamma ray source distribution of a fuel pellet shows Cs migration/vaporization phenomena during irradiation. Using the radial gamma ray source distribution data of Cs-134 and Cs-137, the Cs migration/vaporization time during irradiation can be estimated.

Chemical Isotopic Analyses and Calculations Based on JENDL-3.2 Library for High Burn-up UO2 and MOX Spent Fuels

Chemical isotopic analyses were carried out on high burn-up PWR-UO2, BWR- UO2 and PWR-MOX spent fuels, and ORIGEN2/82 code was verified with the original libraries and new libraries (ORLIBJ32) based on JENDL-3.2. The C/E values evaluated in the literatures were also reviewed. The burn-up and void dependency of the C/E for major actinides and fission products were discussed. The difference of branch ratio from 241Am to 242mAm by (n, γ) reaction between libraries suggests the main reason of underestimation of the C/E of 244Cm with ORLIBJ32.