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Togawa, Orihiko; Okura, Takehisa; Kimura, Masanori
JAEA-Review 2022-049, 76 Pages, 2023/01
JAEA-Review-2022-049.pdf:3.74MB
Before construction and after operation of nuclear facilities, environmental consequence assessments are conducted for normal operation and an emergency. These assessments mainly aim at confirming safety for the public around the facilities and producing relief for them. Environmental consequence assessments are carried out using observations/ measurements by environmental monitoring and/or model predictions by calculation models, sometimes using either of which and at other times using both them, according to the situations and necessities. First, this report investigates methods, roles, merits/demerits and relationship between observations/measurements and model predictions which are used for environmental consequence assessments of nuclear facilities, especially holding up a spent nuclear fuel reprocessing plant at Rokkasho, Aomori as an example. Next, it explains representative examples of utilization of data on observations/measurements and results on model predictions, and considers points of attention at using them. Finally, the report describes future direction, for example, improvements of observations/measurements and model predictions, and fusion of both them.
Yoshida, Kazuo; Tamaki, Hitoshi; Yoshida, Naoki; Amano, Yuki; Abe, Hitoshi
JAEA-Research 2017-015, 18 Pages, 2018/01
JAEA-Research-2017-015.pdf:3.08MB
An accident of evaporation to dryness by boiling of high level liquid waste is postulated as one of the severe accidents at a fuel reprocessing facility. It was observed at the experiments that a large amount of ruthenium (Ru) is volatilized and transfer to the vapor phase in the tank. The nitric acid and water mixed vapor released from the tank is condensed. Volatilized Ru is expected to transfer into the condensed water at the compartments in the building. Quantitative estimation of the amount of Ru transferred condensed water is key issues to evaluate the reduction the amount of Ru through leak path in the facility building. This report presents that a correlation has been developed for Ru transfer rate to condensed water with vapor condensing rate based on the experimental results and additional thermal-hydraulic simulation of the experiments. Applicability of the correlation has been also demonstrated with the accident simulation of typical facilities in full-scale.
Yoshida, Kazuo
JAEA-Research 2016-012, 24 Pages, 2016/08
JAEA-Research-2016-012.pdf:3.04MB
An accident of evaporation to dryness by boiling of high level liquid waste is postulated as one of the severe accidents. In this case, Ru volatilization increases in liquid waste temperature over 120 centigrade at later boiling and dry out phases. It has been observed at the experiments with actual and synthetic liquid waste that some amount of Ru volatilizes and transfers into condensed nitric acid solution at those phases. The nitric acid and water vapor from waste tank condenses at compartments of actual facilities building. The volatilized Ru could transfer into condensed liquid. It is key issues for quantifying the amount of transferred Ru through the facility building to simulate these thermodynamic and chemical behaviors. An analytical model has been proposed in this report based on the condensation mechanisms of nitric acid and water in vapor-liquid equilibria. It has been also carried out to review the thermodynamic properties of nitric acid solution.
Yoshida, Kazuo
JAEA-Research 2016-004, 15 Pages, 2016/06
JAEA-Research-2016-004.pdf:2.22MB
An accident of evaporation to dryness by boiling of high level liquid waste is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, some amount of fission products (FPs) will be transferred to the vapor phase in the tank, and could be released to the environment. Two mitigative accident measures have been proposed by the licensee. One of them is injecting cold water to waste tanks to prevent dryness and another is leading generated vapor through temporary duct to huge spaces in the facility to condense to liquid. Thermal-hydraulics and aerosol transport behaviors in compartments of a typical facility building have been analyzed based on the scenario with these accident measures. The effects of measures are discussed form a view point of the reduction of radioactive material release to environment.
Takeuchi, Masayuki; Yano, Kimihiko; Shibata, Atsuhiro; Sambommatsu, Yuji*; Nakamura, Kazuhito*; Chikazawa, Takahiro*; Hirasawa, Izumi*
Journal of Nuclear Science and Technology, 53(4), p.521 - 528, 2016/04
Times Cited Count：2 Percentile：19.57(Nuclear Science & Technology)Iijima, Shizuka; Uchida, Naoki; Taguchi, Katsuya; Washiya, Tadahiro
JAEA-Review 2015-018, 39 Pages, 2015/11
JAEA-Review-2015-018.pdf:3.95MB
There is a possibility that the fuel assemblies stored in the spent fuel pool (SFP) at Fukushima Daiichi NPS (or Nuclear Power Station) are not only exposed to seawater and concrete fragments, but also damaged by fallen rubbles. We checked the reprocessing experiences of leak fuels at Tokai Reprocessing Plant and overseas reprocessing facilities, and the storage conditions and the checked and inspected results of the fuel stored in the SFP at Fukushima Daiichi NPS, after that, we listed up the technological problems with reprocessing damaged nuclear fuels and selected elements of the research for the purpose of drawing indicators to make a judgmental decision of the possibility of damaged nuclear fuels reprocessing. And we drew the indicators to make a judgmental decision on the possibility of reprocessing based on the results of the examination and the study about elements of the research.
Asakura, Toshihide; Hotoku, Shinobu; Ban, Yasutoshi; Matsumura, Masakazu; Morita, Yasuji
Journal of Nuclear and Radiochemical Sciences, 6(3), p.271 - 274, 2005/12
Tc extraction and separation experiments were performed basing on PUREX technique with using spent UO
fuel with burn-up of 44 GWd/t. The experimental results were examined with performing calculations by a simulation code ESSCAR (Extraction System Simulation Code for Advanced Reprocessing). It was demonstrated that Tc can be almost quantitatively extracted from a dissolver solution and that Tc can also be almost quantitatively recovered by scrubbing. Further, it was clearly presented from the calculation results of ESSCAR that the extraction mechanism of Tc is dominated by the synergistic effect of Zr and U.
Otaki, Kiyoshi*; Tanaka, Yoji*; Katsurai, Kiyomichi*; Aoki, Kazuo*
JAERI-Review 2005-035, 79 Pages, 2005/09
JAERI-Review-2005-035.pdf:4.57MB
In order to collect technical information for the assessment on future nuclear power reactors and fuel cycle systems in Japan, investigation has been made on the characteristics and performance of future reactor options including reduced moderation water reactors (RMWRs) and their fuel cycle systems since the fiscal year 1998. The subjects of investigation are divided into three categories; breeder reactors and their fuel cycle, alternative to sodium-cooled FBR systems,plutonium recycling, spent fuel reprocessing and waste disposal. This report is a summary of the investigation carried out so far.
Morita, Yasuji; Asakura, Toshihide; Mineo, Hideaki; Hotoku, Shinobu; Uchiyama, Gunzo
JAERI-Conf 2005-007, p.25 - 30, 2005/08
Researches on process safety of reprocessing, development of an advanced reprocessing and partitioning of high-level liquid waste(HLLW) have been conducted in NUCEF - BECKY (Back-end Fuel Cycle Key Elements Research Facility), which has alpha-gamma concrete cells and many glove-boxes. This paper presents 10 year accomplishment of the above researches and future activities to be conducted in the field of separation process development.
Asakura, Toshihide; Hotoku, Shinobu; Ban, Yasutoshi; Matsumura, Masakazu*; Kim, S.-Y.; Mineo, Hideaki; Morita, Yasuji
Proceedings of International Conference ATALANTE 2004 Advances for Future Nuclear Fuel Cycles, 5 Pages, 2004/06
In JAERI, PARC process based on PUREX technique has been studied to as the basis of future reprocessing. The key of concept is to obtain the products, U and Pu, within only a single extraction cycle by separating Np and Tc from U and Pu before U/Pu partition. Two flow-sheet tests on the process were performed with 44 GWd/t PWR spent-fuel solutions. It was demonstrated that remaining Np in raffinate from co-extraction could be decreased to 13 % compared to the dissolver solution with increased solvent flow rate and with increased nitric acid concentration of FP scrubbing solution. It was demonstrated that Np separation (selective reduction by n-butyraldehyde) efficiency could be improved from 36 % to 78 % by flow-sheet modification; increasing reductant concentration and scrubbing solution flow rate. The feasibility of the Tc separation technique by high acid scrubbing was demonstrated.
Mineo, Hideaki; Isogai, Hikaru; Morita, Yasuji; Uchiyama, Gunzo*
Journal of Nuclear Science and Technology, 41(2), p.126 - 134, 2004/02
Times Cited Count：7 Percentile：45.06(Nuclear Science & Technology)A simple equation was proposed for the dissolution rate of spent LWR fuel, of which the change in the dissolution area was estimated by taking into account of the area of the cracks occurring due to thermal shrinkage of the pellets during irradiation. The applicability of proposed equation was examined using LWR fuel dissolution test results in the present study as well as the results obtained by other workers. The equation showed good agreements with the dissolution test results obtained from spent fuel pellets and pulverized spent fuel. It was indicated that the proposed equation was simple and would be useful for the prediction of dissolution of spent LWR fuels. However, the initial effective dissolution area, the parameter of the equation, was found to depend on the temperature, which could not be explained by the proposed equation. Further studies on the role of other factors affecting dissolution rate, such as nitrous acid, in the dissolution of spent fuel was required.
Mineo, Hideaki; Goto, Minoru; Iizuka, Masaru*; Fujisaki, Susumu; Hagiya, Hiromichi*; Uchiyama, Gunzo
Separation Science and Technology, 38(9), p.1981 - 2001, 2003/05
Times Cited Count：22 Percentile：63.65(Chemistry, Multidisciplinary)no abstracts in English
Hotoku, Shinobu; Asakura, Toshihide; Mineo, Hideaki; Uchiyama, Gunzo
Journal of Nuclear Science and Technology, 39(Suppl.3), p.313 - 316, 2002/11
no abstracts in English
Tachimori, Shoichi
JAERI-Research 2001-048, 23 Pages, 2001/10
JAERI-Research-2001-048.pdf:1.88MB
A new chemical process, ARTIST process, is proposed for the treatment of spent nuclear fuel. The main concept of the ARTIST process is to recover and stock all actinides (Ans) in two groups, uranium (U) and a mixture of transuranics (TRU), to preserve their resource value and to dispose solely fission products (FPs). The process composed of two main steps, an U exclusive isolation and a total recovery of TRU; which copes with the nuclear non-proliferation measures, and additionally Pu separation process and soft N-donor process if requested, and optionally processes for separation of long-lived FPs. These An products: U-product and TRU-product, are to be solidified by calcination and allowed to the interim stockpile for future utilization. These separations are achieved by use of amidic extractants in accord with the CHON principle. The technical feasibility of the ARTIST process was explained by the performance of both the branched-alkyl monoamides the diglycolic amide (TODGA) in thorough extraction of all TRU by tridentate fashon.
Sawa, Kazuhiro; Fujikawa, Seigo; Yoshimuta, Shigeharu*; Kato, Shigeru*
JAERI-Research 2001-034, 20 Pages, 2001/05
JAERI-Research-2001-034.pdf:1.68MB
no abstracts in English
Mineo, Hideaki; Uchiyama, Gunzo; Hotoku, Shinobu; Asakura, Toshihide; Kihara, Takehiro; *; ; Kimura, Shigeru; *; ; et al.
Proc. of Int. Conf. on Future Nuclear Systems (GLOBAL'99)(CD-ROM), 7 Pages, 1999/00
no abstracts in English
; Nomura, Yasushi; Suyama, Kenya
Proc. of Int. Conf. on the Phys. of Nucl. Sci. and Technol., 1, p.742 - 748, 1998/00
no abstracts in English
Togawa, Orihiko
Journal of Nuclear Science and Technology, 33(10), p.792 - 803, 1996/10
Times Cited Count：3 Percentile：32.75(Nuclear Science & Technology)no abstracts in English
Fujine, Sachio; Takano, Hideki; Yamada, Takashi*; Kurosawa, Katsutoshi*
Global 93 Int. Conf. on Future Nuclear Systems:Emerging Fuel Cycles & Waste Disposal Options, 0, p.1157 - 1162, 1993/00
no abstracts in English
Uchiyama, Gunzo; *; ; *; Sugikawa, Susumu; Maeda, Mitsuru; Tsujino, Takeshi
Radioact. Waste Manage. Nucl. Fuel Cycle, 17(1), p.63 - 79, 1992/00
no abstracts in English
