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Yanagisawa, Hiroshi; Umeda, Miki; Motome, Yuiko; Murao, Hiroyuki
JAEA-Technology 2022-030, 80 Pages, 2023/02
Nuclear criticality benchmark analyses were carried out for TRIGA-type reactor systems in which uranium-zirconium hydride fuel rods are loaded by using the continuous-energy Monte Carlo code MVP with the evaluated nuclear data library JENDL-5. The analyses cover two sorts of benchmark data, the IEU-COMP-THERM-003 and IEU-COMP-THERM-013 in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook, and effective neutron multiplication factors, reactivity worths for control rods etc. were calculated by JENDL-5 in comparison with those by the previous version of JENDL. As the results, it was confirmed that the effective neutron multiplication factors obtained by JENDL-5 were 0.4 to 0.6% greater than those by JENDL-4.0, and that there were no significant differences in the calculated reactivity worths by between JENDL-5 and JENDL-4.0. Those results are considered to be helpful for the confirmation of calculation accuracy in the analyses on NSRR control rod worths, which are planned in the future.
Tonoike, Kotaro; Watanabe, Tomoaki; Gunji, Satoshi; Yamane, Yuichi; Nagaya, Yasunobu; Umeda, Miki; Izawa, Kazuhiko; Ogawa, Kazuhiko
Proceedings of 11th International Conference on Nuclear Criticality Safety (ICNC 2019) (Internet), 9 Pages, 2019/09
Criticality control of the fuel debris in the Fukushima Daiichi Nuclear Power Station would be a risk-informed control to mitigate consequences of criticality events, instead of a deterministic control to prevent such events. The Nuclear Regulation Authority of Japan has administrated a research and development program to tackle this challenge since 2014. The Nuclear Safety Research Center of Japan Atomic Energy Agency, commissioned by the authority, is conducting activities such as computations of criticality characteristics of the fuel debris, development of a criticality analysis code, preparation of criticality experiments, and development of a criticality risk analysis method.
Ozawa, Mayumi; Fukaya, Hiroyuki; Sato, Makoto; Kamohara, Keiko*; Suyama, Kenya; Tonoike, Kotaro; Oki, Keiichi; Umeda, Miki
Proceedings of 53rd Annual Meeting of Hot Laboratories and Remote Handling Working Group (HOTLAB 2016) (Internet), 9 Pages, 2016/11
Tonoike, Kotaro; Yamane, Yuichi; Umeda, Miki; Izawa, Kazuhiko; Sono, Hiroki
Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.20 - 27, 2015/09
From the viewpoint of safety regulation, criticality control of the fuel debris in the Fukushima Daiichi Nuclear Power Station would be a risk-informed control to mitigate consequences of criticality events, instead of a deterministic control to prevent such events. The Nuclear Regulation Authority of Japan has set up a research and development program to tackle this challenge. The Nuclear Safety Research Center of Japan Atomic Energy Agency, commissioned by the authority, has launched activities such as computations of criticality characteristics of the fuel debris, development of criticality risk assessment method, and preparation of criticality experiments to support them.
Suyama, Kenya; Uchiyama, Gunzo; Fukaya, Hiroyuki; Umeda, Miki; Yamamoto, Toru*; Suzuki, Motomu*
Nuclear Back-end and Transmutation Technology for Waste Disposal, p.47 - 56, 2015/00
In fission products in used nuclear fuel, there are several stable isotopes which have large neutron absorption effect. It is known that there are several hardly measurable elements in such important fission products. JAEA had been developed the method to assess the amount of fission products which are hardly measurable and have large neutron capture cross section, under the auspices of the JNES. In this development, the measurement method was developed combining a simple and effective chemical separation scheme of fission products from used nuclear fuel and ICP-MS with high-sensitivity and high-precision. This method was applied to the measurement program for used BWR 99 fuel assembly. This method is applicable to the required measurement for the countermeasure to the accident of the Fukushima Dai-ichi Nuclear Power Plants of Tokyo Electric Power Company. This presentation describes the measurement method developed in the study as well as the future measurement plan in JAEA.
Tonoike, Kotaro; Sono, Hiroki; Umeda, Miki; Yamane, Yuichi; Kugo, Teruhiko; Suyama, Kenya
Nuclear Back-end and Transmutation Technology for Waste Disposal, p.251 - 259, 2015/00
In the Three Mile Island Unit 2 reactor accident, a large amount of fuel debris was formed whose criticality condition is unknown except the possible highest U/U enrichment. The fuel debris had to be cooled and shielded by water in which the minimum critical mass is much smaller than the total mass of fuel debris. To overcome this uncertain situation, the coolant water was borated with sufficient concentration to secure the subcritical condition. The situation is more severe in the damaged reactors of Fukushima Daiichi Nuclear Power Station, where the coolant water flow is practically "once through". Boron must be endlessly added to the water to secure the subcritical condition of the fuel debris, which is not feasible. The water is not borated relying on the circumstantial evidence that the xenon gas monitoring in the containment vessels does not show a sign of criticality. The criticality condition of fuel debris may worsen due to the gradual drop of its temperature, or the change of its geometry by aftershocks or the retrieval work, that may lead the criticality. To avoid criticality and its severe consequences, a certain principle of criticality control must be established. There may be options, such as prevention of the criticality by coolant water boration or by neutronic monitoring, prevention of the severe consequences by intervention measures against criticality, etc. Every option has merits and demerits that must be adequately evaluated toward selection of the best principle.
Kobayashi, Fuyumi; Ishii, Junichi; Shirahashi, Koichi; Umeda, Miki; Sakuraba, Koichi
Kakuhan, Kongo Gijutsu To Toraburu Taisaku, p.341 - 344, 2014/10
The silver mediated electrochemical oxidation (Ag/MEO) process with the ultrasound agitation has been developed for the purpose of the mineralization of organic wastes containing transuranium nuclides at the nuclear fuel reprocessing process. In the Ag/MEO process, organic solvents are decomposed by divalent silver cations under the relatively low temperature and the ambient pressure condition. The ultrasound agitation is effective in mixing the electrolytic solutions and the organic solvents, and is expected to promote the oxidation of the organic solvents. Therefore, the Ag/MEO process with the ultrasound agitation could be a candidate for the treatment of organic solvents. Destruction tests of TBP and dodecane by the Ag/MEO process were conducted to optimize some treatment conditions. Under optimized conditions, the destruction tests of kerosene and TODGA were carried out. It was confirmed that the Ag/MEO process is effective for the mineralization of these organic solvents.
Tonoike, Kotaro; Izawa, Kazuhiko; Sono, Hiroki; Umeda, Miki; Yamane, Yuichi
Transactions of the American Nuclear Society, 110(1), p.282 - 285, 2014/06
no abstracts in English
Fukaya, Hiroyuki; Suyama, Kenya; Sonoda, Takashi; Okubo, Kiyoshi; Umeda, Miki; Uchiyama, Gunzo
JAEA-Research 2013-020, 81 Pages, 2013/10
Japan Atomic Energy Agency conducted a project "Isotopic Composition measurement of Fission Products in Spent Fuel from FY2008 to FY2011" by the entrustment of Japan Nuclear Energy Safety Organization. In that project, we measured the isotopic composition of neodymium isotopes which are important to evaluate the burnup value of spent nuclear fuel by using two different methods and obtained different results. So that we carried out the follow-up measurement in order to investigate the reason of the difference between two neodymium measurements. It was found that we needed correction to the measurement results of neodymium for two samples and a part of other fission products for all samples in total five samples. This report summarizes the all works carried out in this follow-up measurement and obtained results.
Tonoike, Kotaro; Sono, Hiroki; Umeda, Miki; Yamane, Yuichi; Kugo, Teruhiko; Suyama, Kenya
Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.729 - 735, 2013/09
JAEA is conducting studies on criticality control of the fuel debris formed in the accident of Fukushima-Daiichi site. A new control principle must be established, referring principles for existing facilities, and based on criticality characteristics of the debris. In accordance with the principle, safe and practical control has to be realized for the debris whose condition is uncertain at present. This report outlines the present condition of debris and Fukushima site, introduces examples of criticality analysis, and discusses control principles. Research subjects are also proposed to realize the control.
Kokusen, Junya; Sumiya, Masato; Seki, Masakazu; Kobayashi, Fuyumi; Ishii, Junichi; Umeda, Miki
JAEA-Technology 2012-041, 32 Pages, 2013/02
Uranyl nitrate solution fuel used in the STACY and the TRACY is adjusted in the Fuel Treatment System, in which such parameters are varied as concentration of uranium, free nitric acid, soluble neutron poison, and so on. Operations for concentration and denitration of the solution fuel were carried out with an evaporator from JFY 2004 to JFY 2008 in order to adjust the fuel to the experimental condition of the STACY and the TRACY. In parallel, the solution fuel in which some kinds of soluble neutron poison were doped was also adjusted in JFY 2005 and JFY 2006 for the purpose of the STACY experiments to determine neutron absorption effects brought by fission products, etc. After these experiments in the STACY, a part of the solution fuel including the soluble neutron poison was purified by the solvent extraction method with mixer-settlers in JFY 2006 and JFY 2007. This report summarizes operation data of the Fuel Treatment System from JFY 2004 to JFY 2008.
Umeda, Miki; Sugiyama, Tomoyuki; Nagase, Fumihisa; Fuketa, Toyoshi; Ueta, Shohei; Sawa, Kazuhiro
Journal of Nuclear Science and Technology, 47(11), p.991 - 997, 2010/11
Times Cited Count:12 Percentile:67.92(Nuclear Science & Technology)Ishii, Junichi; Kobayashi, Fuyumi; Uchida, Shoji; Sumiya, Masato; Kida, Takashi; Shirahashi, Koichi; Umeda, Miki; Sakuraba, Koichi
JAEA-Technology 2009-068, 20 Pages, 2010/03
At Nuclear Fuel Cycle Safety Engineering Research Facility, the cerium mediated electrolytic oxidation method which is a decontamination technique to decrease the radioactivity of TRU wastes to the clearance-level has been developed for the effective reduction of TRU wastes generated from the decommissioning of a nuclear fuel reprocessing facility and so on. This method corrodes the oxide layer and the surface of metallic TRU metal wastes by the strong oxidation power of Ce in nitric acid. In this study, parameter tests were conducted to optimize the solution condition of Ce initial concentrations and nitric acid concentrations. The target corrosion rate of metallic TRU wastes set to be 24m/h for the practical use of this method. Under the optimized solution condition, a dissolution test of stainless steel simulating wastes was carried out. From the result of the dissolution test, the average corrosion rate was 3.3 m/h during the test time of 90 hours. Based on the supposition that the corrosion depth of metallic TRU wastes was 20 m enough to achieve the clearance-level, the treatment time for the decontamination was about 6 hours. It was confirmed from the result that the decontamination could be performed within one day and the decontamination solution could repeatedly reuse 15 times.
Kobayashi, Fuyumi; Ishii, Junichi; Shirahashi, Koichi; Umeda, Miki; Sakuraba, Koichi
JAEA-Technology 2009-056, 16 Pages, 2009/11
The silver mediated electrochemical oxidation (Ag/MEO) process with the ultrasound agitation has been developed for the purpose of the mineralization of organic wastes containing transuranium nuclides at the nuclear fuel reprocessing process. In the Ag/MEO process, organic solvents are decomposed by divalent silver cations under the relatively low temperature and the ambient pressure condition. The ultrasound agitation is effective in mixing the electrolytic solutions and the organic solvents, and is expected to promote the oxidation of the organic solvents. Therefore, the Ag/MEO process with the ultrasound agitation could be a candidate for the treatment of organic solvents. Destruction tests of TBP and dodecane by the Ag/MEO process were conducted to optimize some treatment conditions. Under optimized conditions, the destruction tests of kerosene and TODGA were carried out. It was confirmed that the Ag/MEO process is effective for the mineralization of these organic solvents.
Sugiyama, Tomoyuki; Umeda, Miki; Udagawa, Yutaka; Sasajima, Hideo; Suzuki, Motoe; Fuketa, Toyoshi
Proceedings of OECD/NEA Workshop on Nuclear Fuel Behaviour during Reactivity Initiated Accidents (CD-ROM), 12 Pages, 2009/09
Sugiyama, Tomoyuki; Umeda, Miki; Sasajima, Hideo; Suzuki, Motoe; Fuketa, Toyoshi
Proceedings of Top Fuel 2009 (DVD-ROM), p.489 - 496, 2009/09
Fuketa, Toyoshi; Sugiyama, Tomoyuki; Umeda, Miki; Sasajima, Hideo; Nagase, Fumihisa
Proceedings of Top Fuel 2009 (DVD-ROM), p.465 - 472, 2009/09
Sugiyama, Tomoyuki; Umeda, Miki; Fuketa, Toyoshi; Sasajima, Hideo; Udagawa, Yutaka; Nagase, Fumihisa
Annals of Nuclear Energy, 36(3), p.380 - 385, 2009/04
Times Cited Count:20 Percentile:79.26(Nuclear Science & Technology)Pulse irradiation tests of high burnup fuels were performed to assess the fuel failure limit in a postulated reactivity-initiated accident (RIA). A BWR-UO rod at a burnup of 69 GWd/t failed due to pellet-cladding mechanical interaction (PCMI). The fuel enthalpy at failure was close to those for PWR-UO rods of 71 to 77 GWd/t with more corroded cladding. Comparison of cladding metallographs between the BWR and PWR fuels showed the morphology of hydride precipitation, which depends on the cladding texture, affects the failure limit. Two tests with PWR-MOX rods of 48 and 59 GWd/t also resulted in PCMI failure. The fuel enthalpies at failure were consistent with results obtained in the previous tests with UO fuel rods, if the failure enthalpy is plotted as a function of the cladding oxide thickness. Therefore, the PCMI failure limit under RIA conditions depends on the cladding corrosion states, and the same limit is applicable to UO and MOX fuels below 59 GWd/t.
Sugiyama, Tomoyuki; Umeda, Miki; Fuketa, Toyoshi; Sasajima, Hideo; Udagawa, Yutaka; Nagase, Fumihisa
Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 8 Pages, 2008/09
Pulse irradiation tests of high burnup fuels were performed to assess the fuel failure limit in a postulated reactivity-initiated accident (RIA). A BWR-UO rod at a burnup of 69 GWd/t failed due to pellet-cladding mechanical interaction (PCMI). The fuel enthalpy at failure was close to those for PWR-UO rods of 71 to 77 GWd/t with more corroded cladding. Comparison of cladding metallographs between the BWR and PWR fuels showed the morphology of hydride precipitation, which depends on the cladding texture, affects the failure limit. Two tests with PWR-MOX rods of 48 and 59 GWd/t also resulted in PCMI failure. The fuel enthalpies at failure were consistent with results obtained in the previous tests with UO fuel rods, if the failure enthalpy is plotted as a function of the cladding oxide thickness. Therefore, the PCMI failure limit under RIA conditions depends on the cladding corrosion states, and the same limit is applicable to UO and MOX fuels below 59 GWd/t.
Umeda, Miki; Ueta, Shohei; Sugiyama, Tomoyuki
Transactions of the American Nuclear Society, 98(1), P. 987, 2008/06