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Tada, Kenichi; Sakino, Takao*
Proceedings of 11th International Conference on Nuclear Criticality Safety (ICNC 2019) (Internet), 9 Pages, 2019/09
Criticality safety of the fuel debris is one of the most important issues, and the adoption of burnup credit is desired. To adopt the burnup credit, validation of the burnup calculation codes is required. In this study, assay data of the used nuclear fuel (2F2DN23, 2F1ZN2, and 2F1ZN3) are evaluated to validate the SWAT4.0 code. The calculation results revealed that the number densities of many heavy nuclides and fission products show good agreement with the experimental data. To investigate the applicability of SWAT4.0 to the criticality safety evaluation of fuel debris, we evaluated the effect of isotopic composition difference on 
. The differences in the number densities of U-235, Pu-239, Pu-241, and Sm-149 have a large impact on . However, the reactivity uncertainty related to the burnup analysis was less than 3%. SWAT4.0 appropriately analyses the isotopic composition of BWR fuel, and it has sufficient accuracy to be adopted in the burnup credit evaluation of fuel debris.
Kikuchi, Takeo; Tada, Kenichi; Sakino, Takao; Suyama, Kenya
JAEA-Research 2017-021, 56 Pages, 2018/03
JAEA-Research-2017-021.pdf:2.15MBJAEA-Research-2017-021(errata).pdf:0.13MB
The criticality management of the fuel debris is one of the most important research issues in Japan. The current criticality management adopts the fresh fuel assumption. The adoption of the fresh fuel assumption for the criticality control of the fuel debris is difficult because the k
of the fuel debris could exceed 1.0 in most of cases which the fuel debris contains water and does not contain neutron absorbers such as gadolinium. Therefore, the adoption of the burnup credit is considered. The prediction accuracy of the isotopic composition of used nuclear fuel must be required to adopt the burnup credit for the treatment of the fuel debris. JAEA developed a burnup calculation code SWAT4.0 to obtain reference calculation results of the isotopic composition of the used nuclear fuel. This code is used to evaluate the composition of fuel debris. In order to investigate the prediction accuracy of SWAT4.0, we analyzed the PIE of BWR obtained from 2F2DN23.
Tada, Kenichi; Kikuchi, Takeo*; Sakino, Takao; Suyama, Kenya
Journal of Nuclear Science and Technology, 55(2), p.138 - 150, 2018/02
Times Cited Count：3 Percentile：29.78(Nuclear Science & Technology)The criticality safety of the fuel debris in Fukushima Daiichi Nuclear Power Plant is one of the most important issues and the adoption of the burnup credit is desired for the criticality analysis. The assay data of used nuclear fuel irradiated in 2F2 is evaluated to validate SWAT4.0 for BWR fuel burnup problem. The calculation results revealed that number density of many heavy nuclides and FPs showed good agreement with the experimental data except for 
U, 
Np, 
Pu and Sm isotopes. The cause of the difference is assumption of the initial number density and void ratio and overestimation of the capture cross section of Np. The C/E-1 values do not depend on the types of fuel rods (UO
or UO-GdO
) and it is similar to that for the PWR fuel. These results indicate that SWAT4.0 appropriately analyzes the isotopic composition of the BWR fuel and it has sufficient accuracy to be adopted in the burnup credit evaluation of the fuel debris.
Kikuchi, Takeo; Tada, Kenichi; Suyama, Kenya
no journal, ,
To estimate the prediction accuracy of the integrated burn up analysis code system SWAT4, we compared the calculation results of SWAT4 and the PIE data of the BWR fuel which was measured by JAERI in 1990s. Comparison results are indicated that the C/E value of major heavy nuclei, e.g., U and Pu, is approximately 1.0. The calculation results are also indicated that some fission products, e.g., Sm, have the larger difference.
Tada, Kenichi; Suyama, Kenya
no journal, ,
To investigate the applicability of the burnup calculation code SWAT4.0, we calculated the post irradiation examination which is 8
8 BWR fuel assembly and loaded in Fukushima Daini Nuclear Power Plant unit 2. In this study, we investigate the impact of the difference of the number density of nuclide on the criticality analysis for the fuel debris. The calculation was carried out by the infinite pin-cell geometry and calculation results indicates that the impact of the difference of the number density of nuclide on the k-effective is less than 3%
k/k. Since we adopt the upper subcriticality limit of 0.95 in usual criticality safety evaluation, this result indicates that SWAT4.0 has adequate prediction accuracy of the isotopic composition to be adopted in the burnup credit analysis of the fuel debris from the Fukushima Daiichi Nuclear Power Plant.
Watanabe, Tomoaki; Kikuchi, Takeo; Tonoike, Kotaro; Suyama, Kenya
no journal, ,
In order to obtain new post irradiation examination data for validation of burnup calculation codes, isotopic compositions of 9 samples obtained from a 1717 high-burnup lead use assembly (53 GWd/t), which was irradiated in Ohi Nuclear Power Station Unit 4, were measured. The isotopic compositions were also evaluated by burnup calculations with SWAT4.0 code and compared with the measurements.
Watanabe, Tomoaki; Kikuchi, Takeo; Kasuya, Yuta*; Nomura, Takuro*; Suyama, Kenya
no journal, ,
JAEA has developed and maintained a burnup calculation code system SWAT4 to evaluate the nuclide composition of spent fuel. Due to the validations, various post-irradiation examinations (PIE) analyses have been performed using the JENDL-4.0 cross-section data. In this study, we performed the previous PIE analyses using the newly released JENDL-5 cross section data to confirm the changes in the analysis results due to the change to JENDL-5 and the differences in the results due to the reactor system and fuel type of the targeted PIE data. As a result of the analysis, it was found that the C/E values of nuclide composition by JENDL-5 did not show any significant difference in trend by reactor system or fuel type, and the results of JENDL-5 and JENDL-4.0 were similar in accuracy for all reactor systems and fuel types. As for the effect of the updated cross sections from JENDL-4.0 to JENDL-5 on C/E values, significant changes in C/E values for Pu-238, Am-241, and Cs-134 were observed due to the updated capture cross sections for Pu-238, Am-241, and Cs-133, independent of reactor system or fuel type existed. The results were highly dependent on the PIE samples as to whether the differences from the measured values were improved or not.
