Ozawa, Takayuki; Hiroka, Shun; Kato, Masato; Novascone, S.*; Medvedev, P.*
Journal of Nuclear Materials, 553, p.153038_1 - 153038_16, 2021/09
To evaluate the O/M dependence of pore migration regarding fuel restructuring at the beginning of irradiation, we are developing BISON for MOX in cooperation with INL and have installed pore migration model considering vapor pressure of vapor species and thermal conductivity for MOX. The O/M dependence of fuel restructuring observed in MA-bearing MOX irradiation experiment in Joyo was evaluated by the 2-dimensional analyses. Four MA-bearing MOX pins with different O/M ratio and pellet/cladding gap size were irradiated in Joyo B14 experiment. Remarkable restructuring of stoichiometric MA-bearing MOX fuels was observed in PIE, and could be evaluated by considering the influence of O/M ratio on vapor pressure. Also, a central void assumes to move toward wide-gap side when the pellet eccentricity taking place, but 2-dimentional analyses on pellet transverse section revealed that the central void formation observed in PIE would be inconsistent with a direction of the pellet eccentricity.
Tsujimura, Norio; Yamazaki, Takumi; Takada, Chie
Journal of Nuclear Science and Technology, 58(1), p.40 - 44, 2021/01
Tanaka, Kosuke; Sato, Isamu*; Onishi, Takashi; Ishikawa, Takashi; Hirosawa, Takashi; Katsuyama, Kozo; Seino, Hiroshi; Ohno, Shuji; Hamada, Hirotsugu; Tokoro, Daishiro*; et al.
Journal of Nuclear Materials, 536, p.152119_1 - 152119_8, 2020/08
In order to obtain the release rate coefficients from fuels for fast reactors (FRs), heating tests and the subsequent analyses of the fission products (FPs) and actinides that are released were carried out using samples of uranium-plutonium mixed oxide (MOX) fuel pellets irradiated at the experimental fast reactor Joyo. Three heating tests targeting temperatures of 2773, 2973 and 3173 K were conducted using an FP release behavior test apparatus equipped with a high-frequency induction furnace and solid FP sampling systems consisting of a thermal gradient tube (TGT) and filters. Irradiated fuel pellets were placed into a tungsten crucible, then loaded into the induction furnace. The temperature was raised continuously at a heating rate of 10 K/s to the targeted temperature and maintained for 500 s in a flowing argon gas atmosphere. The FPs and actinides released from the MOX fuels and deposited in the TGT and filters were quantified by gamma-ray spectrometry and inductively coupled plasma mass spectrometry (ICP-MS) analysis. Based on the analysis, the release rates of radionuclides from MOX fuels for FR were obtained and compared with literature data for light water reactor (LWR) fuels. The release rate coefficients of FPs obtained in this study were found to be similar to or lower than the literature values for LWR fuels. It was also found that the release rate coefficient data for actinides were within the range of variation of literature values for LWR fuels.
Ikusawa, Yoshihisa; Hiroka, Shun; Uno, Masayoshi*
2018 GIF Symposium Proceedings (Internet), p.321 - 327, 2020/05
Research and development of Minor actinides (MAs) bearing MOX fuel for fast reactor has been proceeding from the viewpoint of reducing radioactive waste. In order to develop, MA bearing MOX, it is indispensable to clarify the influence of MA addition on irradiation behavior. The addition of Americium (Am) to MOX affects vapor pressure and thermal conductivity, which are important properties from the perspective of evaluating fuel temperature. This is because vapor pressure affects fuel restructuring, and thermal conductivity affects fuel temperature distribution. Focusing on these physical properties, this study evaluates the influence of Am on fuel temperature using irradiation behavior analysis code to contribute to the development of MA-bearing MOX fuel. An increase in Am content decreases the thermal conductivity and increases the oxygen potential of oxide fuel. Because vapor pressure increases with increasing Am content, pore migration is accelerated, and the central void diameter increases with increasing Am content. As a result, after formation of the central void, the influence of Am content on the fuel center temperature is mild. Alpha particles generated by radioactive decay of transuranium elements cause lattice defects in the oxide fuel pellets. It is well known that this phenomenon, which is called self-irradiation, affects thermal conductivity. Since americium is the typical alpha radioactive nucleus, to evaluate fuel temperature of Am-MOX is necessary to take account of the influence of self-irradiation damage on thermal conductivity. Self-irradiation decreases thermal conductivity, and as the Am content increases, the rate of decrease in thermal conductivity is accelerated. Because it recovers with temperature rise, the decrease in thermal conductivity due to self-irradiation damage has very little effect on fuel center temperature. These results suggest that Am-MOX fuel could be irradiated under the same conditions as conventional MOX fuel.
Ishii, Katsunori; Segawa, Tomomi; Kawaguchi, Koichi; Suzuki, Masahiro
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 5 Pages, 2019/05
Japan Atomic Energy Agency (JAEA) is developing a simplified pelletizing process for MOX fuel fabrication. In this process, the flowability of MOX powder produced by de-nitration conversion based on microwave heating, calcination, and reduction is improved using the wet granulation method. In a previous paper, to produce MOX granules of appropriate sizes for pelletizing them effectively, we proposed a granulation system composed of a wet granulator and a sizing machine. In the present work, we modernized the wet granulator, completed the granulation system by adding auxiliary equipment, and conducted performance tests of the granulation system with WO powder. The results of a performance test indicated that it is possible to convert raw powder into granules characterized by appropriate size and excellent flowability. The time required to process 5 kg of WO powder was about 70 min, which almost satisfies the target time.
Ikusawa, Yoshihisa; Morimoto, Kyoichi; Kato, Masato; Saito, Kosuke; Uno, Masayoshi*
Nuclear Technology, 205(3), p.474 - 485, 2019/03
This study evaluated the effects of plutonium content and self-irradiation on the thermal conductivity of mixed-oxide (MOX) fuel. Samples of UO fuel and various MOX fuels were tested. The MOX fuels had a range of plutonium contents, and some samples were stored for 20 years. The thermal conductivity of these samples was determined from thermal diffusivity measurements taken via laser flash analysis. Although the thermal conductivity decreased with increasing plutonium content, this effect was slight. The effect of self-irradiation was investigated using the stored samples. The reduction in thermal conductivity caused by self-irradiation depended on the plutonium content, its isotopic composition, and storage time. The reduction in thermal conductivity over 20 years' storage can be predicted from the change of lattice parameter. In addition, the decrease in thermal conductivity caused by self-irradiation was recovered with heat treatment, and recovered almost completely at temperatures over 1200 K. From these evaluation results, we formulated an equation for thermal conductivity that is based on the classical phonon-transport model. This equation can predict the thermal conductivity of MOX fuel thermal conductivity by accounting for the influences of plutonium content and self-irradiation.
Kihara, Yoshiyuki; Tanaka, Kosuke; Koyama, Shinichi; Yoshimochi, Hiroshi; Seki, Takayuki; Katsuyama, Kozo
NEA/NSC/R(2017)3, p.341 - 350, 2017/11
In order to investigate the effect of the addition of americium to MOX fuels on the irradiation behaviour, the "Am-1" program is being conducted at the JAEA. The Am-1 program consists of two short-term irradiation tests of 10-min and 24-h irradiation periods, and a steady-state irradiation test. The short-term irradiation tests and their post irradiation examinations (PIEs) have been successfully completed. To date, the data for PIE of the Am-MOX fuels focused on the microstructural evolution and redistribution behaviour of Am at the initial stage of irradiation have been obtained and reported. In this paper, the results obtained from the Am-1 program are reviewed and detailed descriptions of the fabrication and inspection techniques for the Am-MOX fuels prepared for the program are provided. PIE data for the Am-MOX fuels at the initial stage of irradiation have been accumulated. In this paper, unpublished PIE data for the Am-MOX fuels are also presented.
Ikusawa, Yoshihisa; Maeda, Koji; Kato, Masato; Uno, Masayoshi*
Nuclear Technology, 199(1), p.83 - 95, 2017/07
Based on thermal computation results obtained using an irradiation behavior analysis code, we have evaluated the effect of O/M ratio on fuel restructuring from the results of PIEs for the B14 irradiation test fuel, which was a mixed oxide fuel and was irradiated in the experimental reactor Joyo. The thermal computation results showed that fuel restructuring in the stoichiometric oxide fuel was accelerated, though the fuel temperature in the stoichiometric oxide fuel was evaluated as lower than that of the hypo-stoichiometric one. We explained this behavior as follows: first, the fuel temperature decreased due to the high thermal conductivity at stoichiometry; second, the pore migration velocity increased due to the increase in vapor pressure caused by the high vapor pressure of UO, which was derived from the high oxygen potential at stoichiometry. In addition, our results indicated that the central void diameter strongly depended on not only fuel temperature, but also vapor pressure.
Fujita, Shunya*; Abe, Yutaka*; Kaneko, Akiko*; Chonan, Fuminori*; Yuasa, Tomohisa*; Yamaki, Tatsunori*; Segawa, Tomomi; Yamada, Yoshikazu
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 8 Pages, 2017/07
From the observation results, in the process of flushing, the behaviors leading to flushing were classified divided into three types. First type is that first generation bubble from heating leads to flushing. Second type is that nucleate boiling continues during heating and stop, finally single bubble generates and leads to flushing. Third type is defined that gradual evaporation occurs without bubbles. It was revealed that the total quantities of heat released by flushing are approximately equal when assuming the flushing mechanism, it can be triggered that a large amount of micro bubbles are instantaneously generated and grew.
Watanabe, Masashi; Sunaoshi, Takeo*; Kato, Masato
Defect and Diffusion Forum, 375, p.84 - 90, 2017/05
The oxygen chemical diffusion coefficient in (U, Pu)O was determined by thermo-gravimetry as functions of the Pu content, oxygen-to-metal ratio and temperature. The surface reaction was considered in the diffusion coefficient determination. The activation energy for the chemical diffusion coefficient was 60 kJ/mol and 65 kJ/mol, respectively, in (UPu)O and (UPu)O.
Kato, Masato; Watanabe, Masashi; Matsumoto, Taku; Hiroka, Shun; Akashi, Masatoshi
Journal of Nuclear Materials, 487, p.424 - 432, 2017/04
Oxygen potential of (U,Pu)O was evaluated based on defect chemistry using an updated experimental data set. The relationship between oxygen partial pressure and deviation in (U,Pu)O was analyzed, and equilibrium constants of defect formation were determined as functions of Pu content and temperature. Brouwer's diagrams were constructed using the determined equilibrium constants, and a relational equation to determine O/M ratio was derived as functions of O/M ratio, Pu content and temperature. In addition, relationship between oxygen potential and oxygen diffusion coefficients were described.
Tokai Reprocessing Technology Development Center
JAEA-Evaluation 2015-012, 83 Pages, 2015/12
Japan Atomic Energy Agency (hereafter referred as "JAEA") consulted the "Evaluation Committee of Research and Development Activities for Fast Reactor Cycle" to assess the issue on "Research and Development on Reprocessing of Nuclear Fuel Materials" conducted by JAEA during the period from FY2010 to FY2014. In response to the JAEA's request, the committee assessed the R&D programs and the activities of JAEA related to the issue and concluded the mission was accomplished. This evaluation was performed based on the "General guideline for the evaluation of government R&D activities", the "Guideline for evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology (MEXT)" and the "Operational rule for evaluation of R&D activities" by JAEA.
Kimura, Yasuhisa; Hirano, Hiroshi; Watahiki, Masatoshi; Kuba, Meiji; Ishikawa, Shinichiro
Dekomisshoningu Giho, (52), p.45 - 54, 2015/09
The Plutonium Fuel Fabrication Facility (PFFF) of the Japan Atomic Energy Agency is now in its decommissioning phase. In the PFFF, terminated gloveboxes have been dismantled. Gloveboxes to be dismantled are surrounded by a plastic enclosure to prevent contamination from being spread into process room. Dismantling operations for gloveboxes are performed manually by workers, each wearing an air-feed suit. However, the mental and physical loads placed on workers wearing the air-feed suits are intensively high. Therefore, R&Ds on new dismantling technologies including utilization of heavy machines covered with plastic enclosure for anti-contamination have been started to reduce the potential risks associated with workers and decommissioning costs. In this paper, the status of decommissioning of the PFFF and the overview of developed dismantling technologies for -tight gloveboxes are described.
Yamaki, Tatsunori*; Abe, Yutaka*; Kaneko, Akiko*; Segawa, Tomomi; Kawaguchi, Koichi; Yamada, Yoshikazu
Proceedings of the 22nd International Conference on Nuclear Engineering 2014 (ICONE-22), Vol.2A, p.V02AT09A011_1 - V02AT09A011_10, 2014/07
Uranium and plutonium mixed nitrate solution of the spent nuclear fuel is converted to uranium and plutonium mixed oxide (MOX) powder by the microwave heating direct denitration method in co-conversion process of the nuclear fuel cycle. The cylindrical denitration vessel can be expected to realize high-speed and high-capacity processing against traditional shallow vessel. However, flushing and overflow phenomena of solution have been confirmed in cylindrical vessel. Thus, It is required to be clarified that the generation condition and generation mechanism of flushing phenomena during microwave heating for the safety of the operating condition and the optimum design of the equipment. The craiteria of flushing phenomena and the relationship between the released power by evaporation and the absorbed power in water is confirmed by the experimental results from the microwave heating. It is found that the flow structure is changed and the outer surface is heated according to the increase of the concentration of KCl by the microwave heating experiment with using the KCl solution and KCl jelly as a simulated solution of the uranium and plutonium mixed nitrate solution.
Abe, Hitoshi; Tashiro, Shinsuke; Miyoshi, Yoshinori
Nihon Genshiryoku Gakkai Wabun Rombunshi, 6(1), p.10 - 21, 2007/03
In MOX fuel fabrication facility, zinc stearate will be added into the MOX powder as addition material. If the material is added in large excess by miss operation, criticality characteristics of the MOX fuel would be influenced because it has neutron moderation effect. If criticality condition should be induced by the excess addition, physical variations, such as melting and pyrolysis of the material, must be caused by the fission energy and dynamic characteristics of the MOX fuel must be affected. To contribute quantitative evaluation of the dynamic characteristics, thermal properties data such as exo/endothermic calorific values, reaction rates, etc. with the respective physical variations and release behavior of pyrolysis gas were measured. It was found the exo/endothermic behavior with rinsing temperature of the material could be divided into six regions and rapid pressure rise caused by the pyrolysis reaction over about 400 C. Furthermore, on the basis of the results, evaluation model for the thermal properties under the criticality condition was also investigated.
Committee of the Halden Joint Research Programme
JAERI-Tech 2004-023, 38 Pages, 2004/03
JAERI has performed cooperative researches with several Japanese organizations utilizing the Halden Boiling Heavy Water Reactor(HBWR) which is located at Halden in Norway. These researches are carried out based on the contracts of the cooperative researches, which are revised every three years, in accordance with the renewal of the participation of JAERI to the OECD Halden Reactor Project. This report summarizes the objectives, contents and the outlines of the achievements of the cooperative researches during the three years from 2000 January to 2002 December. During the period, seven cooperative researches had been carried out. Two of them had been completed and other five researches have been continued to the next three years period. Most of them are irradiation test researches of advanced fuel and cladding in order to prepare the higher burnup utilization and introduction of LWR fuel and MOX fuel in LWRs of Japan.
Okubo, Tsutomu; Iwamura, Takamichi; Takeda, Renzo*; Yamauchi, Toyoaki*; Okada, Hiroyuki*
Proceedings of International Conference on Global Environment and Advanced Nuclear Power Plants (GENES4/ANP 2003) (CD-ROM), 8 Pages, 2003/00
A water-cooled reactor concept named Reduced-Moderation Water Reactor is under development for effective fuel utilization through plutonium multiple recycling based on the water-cooled reactor technology. The reactor aims at achievement of a high conversion ratio more than 1.0 with MOX fuel. Especially, the core performances during the Pu multiple recycling have been investigated for the advanced fuel reprocessing schemes with low decontamination factors than the current PUREX process, and are shown to achieve the conversion ratio more than 1.0 and the negative void reactivity coefficient.
Nakano, Yoshihiro; Ishikawa, Nobuyuki; Nakatsuka, Toru; Iwamura, Takamichi
JAERI-Conf 2002-012, 219 Pages, 2002/12
no abstracts in English
Okubo, Tsutomu; Suzuki, Motoe; Iwamura, Takamichi; Takeda, Renzo*; Moriya, Kumiaki*; Kanno, Minoru*
Proceedings of International Conference on the New Frontiers of Nuclear Technology; Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002) (CD-ROM), 10 Pages, 2002/10
A small scale around 300 MWe reduced-moderation water reactor (RMWR) concept has been developed. For the core, a BWR type core concept with the tight-lattice fuel rod arrangement and the high void fraction is adopted to attain a high conversion ratio over 1.0. The negative void reactivity coefficients are also required, and the very flat short core concept is adopted to make the natural circulation cooling (NC) possible. The core burn-up of 60 GWd/t and the operation cycle of 24 months are also attained. For the system, simplification of the system with the passive safety features is a basic approach to overcome the scale demerit as well as the NC. For example, the HPCF system is replaced with the passive accumulator system resulting in the expensive emergency DGs reduction. The cost evaluation for concerned NSSS components gives about 20% reduction. Since MOX fuels in the RMWR contains Pu around 30 wt% and is irradiated to a high burn-up, the fuel safety evaluation has been performed and the acceptable results have been obtained from the thermal feasibility point of view.
Okubo, Tsutomu; Iwamura, Takamichi; Yamamoto, Kazuhiko*; Okada, Hiroyuki*
Nihon Kikai Gakkai Dai-8-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.571 - 574, 2002/00
Based on the experienced light water reactor technology, conceptual design studies on advanced water-cooled reactors have been performed. They are named “Reduced-Moderation Water Reactor" (RMWR) with the high conversion ratio more than 1.0 and the negative void reactivity coefficients. Several concepts have been successfully established for them based on the neutronics calculations. Based on these concepts, detailed investigations on such as plutonium multiple recycling and control rod planning have been performed as well as improvement on core performances. Through these detailed core design investigation, the feasibility of those designs has been confirmed step by step.