Study on criticality safety control of fuel debris for validation of methodology applied to the safety regulation

Suyama, Kenya; Ueki, Taro; Gunji, Satoshi; Watanabe, Tomoaki; Araki, Shohei; Fukuda, Kodai; Yamane, Yuichi; Izawa, Kazuhiko; Nagaya, Yasunobu; Kikuchi, Takeo; et al.

Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 6 Pages, 2023/10

To remove and store safely the fuel debris generated by the severe accident of the Fukushima Daiichi Nuclear Power Station in 2011 is one of the most important and challenging topics for decommissioning of the damaged reactors in Fukushima. To validate the adopted method for the evaluation of criticality safety control of the fuel debris through comparison with the experimental data obtained by the criticality experiments, the Nuclear Regulation Authority (NRA) of Japan funds a research and development project which was entrusted to the Nuclear Safety Research Center (NSRC) of Japan Atomic Energy Agency (JAEA) from 2014. In this project, JAEA has been conducting such activities as i) comprehensive computation of the criticality characteristics of the fuel debris and making database (criticality map of the fuel debris), ii) development of new continuous energy Monte Carlo code, iii) evaluation of criticality accident and iv) modification of the critical assembly STACY for the experiments for validation of criticality safety control methodology. After the last ICNC2019, the project has the substantial progress in the modification of STACY which will start officially operation from May 2024 and the development of the Monte Carlo Code "Solomon" suitable for the criticality calculation for materials having spatially random distribution complies with the power spectrum. We present the whole picture of this research and development project and status of each technical topics in the session.

Criticality characteristics of MCCI products possibly produced in reactors of Fukushima Daiichi Nuclear Power Station

Tonoike, Kotaro; Okubo, Kiyoshi; Takada, Tomoyuki*

Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.292 - 300, 2015/09

The damaged Unit 1-3 reactors of the Fukushima Daiichi Nuclear Power Station may contain fuel debris of a significant amount that is in a form of molten-core-concrete-interaction (MCCI) product with porous structure. Such low density MCCI product including fissile material is a great concern for its criticality control, especially under submerged condition, due to its fairly good neutron moderation. This report shows computation results of basic criticality characteristics of the MCCI product, which will facilitate criticality risk assessments during decommissioning of the reactors. The results imply that water bound in concrete may raise the risk from the viewpoints of possibility of criticality events and of effectiveness of mitigation measures such as neutron poison injection into coolant water.

Infinite multiplication factor of low-enriched UO-concrete system

Izawa, Kazuhiko; Uchida, Yuriko; Okubo, Kiyoshi; Totsuka, Masayoshi; Sono, Hiroki; Tonoike, Kotaro

Journal of Nuclear Science and Technology, 49(11), p.1043 - 1047, 2012/11

AA2012-0375.pdf:0.63MB

https://doi.org/10.1080/00223131.2012.730893

Possibility of criticality of fuel debris in a form of UO-concrete mixture is evaluated by calculating infinite multiplication factor () for a study of criticality control on the fuel debris generated through the molten core concrete interaction (MCCI) in a severe accident of a light water reactor (LWR). The infinite multiplication factor can be greater than unity, which means that handling of the mixture is subject to criticality control. This paper shows that concrete have efficient slowing-down capability of neutron and points out the necessity of further investigations on the criticality of low-enriched UO-concrete system for actual handling of fuel debris.

Re-evaluation of assay data of spent nuclear fuel obtained at Japan Atomic Energy Research Institute for validation of burnup calculation code systems

Suyama, Kenya; Murazaki, Minoru*; Okubo, Kiyoshi; Nakahara, Yoshinori*; Uchiyama, Gunzo

Annals of Nuclear Energy, 38(5), p.930 - 941, 2011/05

https://doi.org/10.1016/j.anucene.2011.01.025

The isotopic composition of spent nuclear fuels is vital data for studies on the nuclear fuel cycle and reactor physics. The Japan Atomic Energy Agency (JAEA) has been active in obtaining such data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuels, and some data has already been published. These data have been registered with the international Spent Fuel Isotopic Composition Database (SFCOMPO) and widely used as international benchmarks for burnup calculation codes and libraries. In this paper, postirradiation examination (PIE) data from two fuel assemblies irradiated in the Ohi-1 and Ohi-2 PWRs in Japan are shown. The PIE data from Ohi-2 has already been published. However, these data were not suitable for the benchmarking of calculation codes and libraries because several important specifications and data were not included. This paper summarizes the details of PIE data from Ohi-1 and Ohi-2 and shows in detail the data and specifications required for analyses of isotopic composition. For precise burnup analyses, the burnup values of PIE samples were re-evaluated in this study. These PIE data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. This indicates that the quality of PIE data from Ohi-1 and Ohi-2 PWRs is high, and that these PIE data are suitable for the benchmarking of burnup calculation code systems.

Active reduction of the end effect by local installation of neutron absorbers

Suyama, Kenya; Murazaki, Minoru; Okubo, Kiyoshi; Okuno, Hiroshi

Annals of Nuclear Energy, 35(9), p.1628 - 1635, 2008/09

https://doi.org/10.1016/j.anucene.2008.02.011

In the analysis of the burnup credit, it has been pointed out that the neutron multiplication factor becomes greater if we consider an axial burnup distribution of spent fuel assemblies than the case under an assumption of an average burnup through the fuel assemblies. This phenomenon is called "end effect" and it is one of the main technical issues in the burnup credit study. In this study, the reason why the end effect occurs in the criticality calculation of spent fuel assemblies is discussed by analyses of neutron flux distribution measurement both fixed source and eigenvalue calculations. These calculations show us that the end effect is induced by the solution of neutron balance equation as eigenvalue problem and an actual neutron flux increase occurs only when the neutron multiplication factor is close to unity. Based on the discussion, reducing the end effect actively by local installation of neutron absorbers (LINA) around the end regions of the fuel assemblies are proposed and its effect was confirmed based on the several criticality calculations.

Criticality safety evaluation of damaged burned nuclear fuel; Effect of structural materials

Okubo, Kiyoshi; Suyama, Kenya; Kashima, Takao; Tonoike, Kotaro; Takada, Tomoyuki*

no journal, , 

Criticality safety analysis is necessary for the damaged-fuel handling in the Fukushima Daiichi NPP decommissioning. This presentation show influence of structural materials such as Zry-2, Fe, concrete expected to be present in the damaged fuel. Multiplication factor (kinf) decreases most by replacing moisture, in the damaged fuel, with iron. Replacement of all moisture with Zry-2 gives the same influence as iron, although decrease rate of kinf is lower because of the smaller absorb cross section of Zry-2. Concrete has much less influence due to the neutron moderation by hydrogen contained in concrete, which calls attention on handling of the concrete-fuel mixture. Effect as reflector of the materials is also evaluated.