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Ozawa, Takayuki; Kato, Masato
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.2036 - 2044, 2009/09
Americium, one of MAs, is contained in MOX fuels due to decay of plutonium-241. The radial redistribution of americium has been observed with that of plutonium in the irradiated MOX fuel. The development of a migration model for plutonium and americium redistribution would be important for fuel design because of their influence on thermal properties, i.e. thermal conductivity and melting temperature. In this study, the migration model for plutonium and americium redistribution was developed by taking into account thermal diffusion concurrently with vapor phase transport via pores in the fuel. The computed radial redistribution of plutonium and americium was found to be in good agreement with the results of post-irradiation examinations after the irradiation test for 2% neptunium and 2% americium doped uranium plutonium mixed oxide (U, Np, Pu, Am)O
fuel in JOYO.
Yano, Kimihiko; Nakahara, Masaumi; Nakamura, Masahiro; Shibata, Atsuhiro; Nomura, Kazunori; Nakamura, Kazuhito*; Tayama, Toshimitsu; Washiya, Tadahiro; Chikazawa, Takahiro*; Kikuchi, Toshiaki*; et al.
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.143 - 150, 2009/09
Shibata, Atsuhiro; Kaji, Naoya; Nakahara, Masaumi; Yano, Kimihiko; Tayama, Toshimitsu; Nakamura, Kazuhito; Washiya, Tadahiro; Myochin, Munetaka; Chikazawa, Takahiro*; Kikuchi, Toshiaki*
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.151 - 157, 2009/09
As a part of FaCT project, Japan Atomic Energy Agency has been developing a U crystallization process for advanced aqueous reprocessing technology in collaboration with Mitsubishi Materials Corporation. We have carried out experimental studies and obtained fundamental data. Continuous operation tests were also carried out by an engineering-scale crystallizer to confirm productivity of the equipment and to investigate non-steady state conditions. The requirements for the U crystallization process in the FaCT project could be achieved except DF of Cs. More detail investigation is under way to settle the process condition without Pu-Cs double salt formation.
Kato, Masato; Segawa, Tomoomi; Takeuchi, Kentaro; Kashimura, Motoaki; Kihara, Yoshiyuki
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.2051 - 2058, 2009/09
In the Fast Reactor Cycle technology Development (FaCT) project conducted by Japan Atomic Energy Agency, minor actinide-containing MOX (MA-MOX) fuel has been developed. The fuel is a homogeneous MOX fuel which contains a maximum of 5% MA such as Am and Np. The oxygen-to-metal (O/M) ratio of the fuel is adjusted to less than 1.97 to control fuel and cladding chemical interaction in a high burn-up of 150 GWd/t. In this paper, the thermal properties of raw powder and sintering behavior of oxidized powder were investigated. In addition the O/M ratio adjustment procedure was established.
Kitawaki, Shinichi; Nakayoshi, Akira; Fukushima, Mineo; Koizumi, Tsutomu; Kurata, Masaki*; Yahagi, Noboru*
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.1269 - 1273, 2009/09
JAEA is developing the pyroreprocessing by collaboration with CRIEPI. The test using U began in 2002, and the test using PuO
and unirradiated MOX were ended in 2008. The reduction of UO pellets by using Li-reduction method, the electrowinning using reduced pellets, the separation of adhered salt with deposit by distillation, and the ingot formation of deposit were performed. As a result, 99% of the loaded U is recovered as metal ingot. The tests similar to U tests were performed by using PuO. As a result, Pu was successfully recovered with U metal. In the MOX test, the mass balance of Pu was maintained at 
100% with respect to the initial amount. We try to form the U-Pu-Zr alloy by using reduced MOX. After 2009, the process development that uses the alloy will be continued.
Nakatsuka, Toru; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.1690 - 1693, 2009/09
In order to assess the thermal hydraulics in fuel assemblies of the supercritical water cooled reactor (SCWR) core, JAEA has been enhancing the three-dimensional two-fluid model analysis code ACE-3D to predict thermal-hydraulic behavior of the SCWR. As a part of these assessments, the present paper describes numerical analysis results on thermal-hydraulic characteristics in the fuel assembly based on the design of the supercritical water cooled fast reactor (Super Fast Reactor) concept. The analytical geometry simulates a 19-rod assembly, which is one of the most simplified geometry of the SCWR fuel assembly and includes three kinds of different subchannel types: (1), adjacent to the channel box; (2), next to type (1); and (3), located inside types (1) and (2). The results show the influence of existence of the channel box and variation of the rod surface temperature profiles in the circumferential direction.
Kondo, Yoshikazu; Uchida, Naoki; Terunuma, Hirotaka; Tanaka, Kosuke; Oyama, Koichi; Katsurai, Kiyomichi; Washiya, Tadahiro
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.277 - 280, 2009/09
The composition of sludge in the dissolver after dissolution of PWR and ATR fuels at the Tokai Reprocessing Plant (TRP) was analyzed. As a result the presence of zirconium molybdate was confirmed by the analysis of X-ray diffraction (XRD). To clarify the formation behavior of the precipitates of zirconium molybdate, investigated the dependence of HNO
concentration on the precipitation with Mo and Zr solution. To evaluate the adhesion on the metal surface (stainless steel and Ti metal), the deposition amounts of the precipitates of zirconium molybdate on the metal were also examined. In addition, it reports on the comparative result of executing a chemical dissolution of the precipitates by using the solutions of NaOH, CHO
-HNO and HO-HNO.
Sugawara, Takanori; Nishihara, Kenji; Sasa, Toshinobu; Oigawa, Hiroyuki
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.1782 - 1790, 2009/09
In Japan Atomic Energy Agency (JAEA), various R&D for 800 MWt, lead bismuth eutectic (LBE) cooled ADS have been performed to transmute minor actinides (MA) discharged from spent fuel of commercial nuclear power plants. This study introduces the latest two activities for the ADS in JAEA. The first one is the uncertainty analysis for the subcritical core. Since uncertainties of the current MA nuclear data are supposed to be larger than those of other nuclides, it is necessary to improve the accuracy of the neutronics design. We discuss the effect of MA-loaded experiments for its improvement by using the cross-section adjustment procedure. The second activity is the structural analysis of the beam window. The discussions for the feasible concept are performed by using Finite Element Method.
Tanai, Kenji; Naito, Morimasa
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.796 - 805, 2009/09
Samoto, Hirotaka; Kimura, Norimichi; Otani, Takehisa; Sugai, Eiji; Hayashi, Shinichiro
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.458 - 463, 2009/09
JAEA has been developing technology to immobilize radioactive krypton gas to metal alloy by ion-implantation method as a stable storage technique of krypton gas recovered from a reprocessing plant. The characteristics of implantation of krypton gas and of krypton implanted alloy were investigated by the cold test performed so far. In this paper, we report the results of the hot immobilization test performed at the Krypton Recovery development Facility (hereafter called KRF) which is attached to the Tokai Reprocessing Plant (hereafter called TRP). In this test, we immobilized the radioactive krypton gas recovered from TRP by cryogenic distillation process of KRF and investigated the gas retention characteristics of the implanted alloy.
Hayashi, Hirokazu; Shibata, Hiroki; Akabori, Mitsuo; Arai, Yasuo; Minato, Kazuo
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.1166 - 1173, 2009/09
R&D on the transmutation of long-lived minor actinides (MA) by the accelerator-driven system (ADS) using nitride fuels is underway at JAEA. In regard to reprocessing technology, pyrochemical process has several advantages in case of treating spent fuel with large decay heat and fast neutron emission, and recovering highly enriched 
N. In the pyrochemical reprocessing, plutonium and MA are dissolved in LiCl-KCl eutectic melts and selectively recovered into liquid cadmium (Cd) cathode by molten salt electrorefining. The electrochemical behavior in LiCl-KCl eutectic melts and the subsequent nitride formation behavior of plutonium and MA recovered in liquid Cd cathode are investigated. Electrochemical study of americium (Am) on electrolyses of AmN in LiCl-KCl eutectic melts and nitride formation of Am recovered in the liquid Cd cathode are presented. Electrochemical behavior of Am on anodic dissolution of AmN and recovery of Am into a liquid Cd cathode by electrolyses in LiCl-KCl eutectic melts was investigated by transient electrochemical techniques. Am was recovered as Am-Cd alloy in the liquid Cd cathode, in which AmCd
type phase was identified besides Cd phase. The recovered Am was converted to AmN by the nitridation-distillation combined method. These results suggest that the pyrochemical process developed for the nitride fuel cycle is applicable for the nitride fuels containing Am, which is to be used for the transmutation of MA.
Tanaka, Kenya; Ishii, Tetsuya; Yoshimochi, Hiroshi; Asaka, Takeo
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.2045 - 2050, 2009/09
As a part of the economic evaluation of the MA recycling system, the management cost of high level radioactive waste was estimated quantitatively. The development of an innovative fuel fabrication process has been done by using UO powder, U metal particles and Mo powder. From comparisons of granulated material characteristics, two candidate methods, mixing granulation (MIX/G) and extruding granulation (EXT/G), were considered to have good feasibility as the fuel fabrication process. In the preliminary sintering test of granulated UO obtained by EXT/G, a high density UO pellet (97% of TD) with 5wt% of U and 5wt% of Mo was successfully sintered. From the results of thermal conductivity measurements, it was confirmed that the dispersion of Mo powder and U metal into the oxide matrix was an effective way to improve the characteristic.
Sato, Takumi; Iwai, Takashi; Arai, Yasuo
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.1278 - 1286, 2009/09
no abstracts in English
Nakamura, Kinya*; Ogata, Takanari*; Kato, Tetsuya*; Nakajima, Kunihisa; Arai, Yasuo
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.1487 - 1495, 2009/09
The U-Pu-Zr fuel slugs for the irradiation test in JOYO were manufactured in a small-scale injection casting furnace. The U-Pu alloy ingots as starting materials were prepared by means of electrochemical reduction of the dioxides. The U-Pu-Zr fuel slugs manufactured met all the specifications determined based on not only the results of preliminary tests but also the specification of EBR-II driver fuels. The americium to plutonium ratio in the fuel slugs slightly decreased after the injection casting process.
Kurita, Tsutomu; Kato, Yoshiyuki; Yoshimoto, Katsunobu; Suzuki, Masahiro; Kihara, Yoshiyuki; Fujii, Kanichi
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.94 - 102, 2009/09
With regard to advanced MOX fuel fabrication, a new concept in which one vessel especially designed to meet microwave de-nitration is utilized also for crushing and for granulation, without organic lubricant nor powder transfer across the processes, was introduced for innovative MOX powder production. In order to realize this concept, two attempts were made: A specially designed three blade impeller coupled with auxiliary blade. A uniquely shaped mixing blade coupled with an auxiliary blade having auto-orbital hybrid rotation. The mixing blade promotes the growth of particles, whereas the auxiliary blade suppresses the overgrowth by chopping larger particles. These granulators use a little water as binder. As a result, major diameter of granule 400-1000 micron and flow-ability 82-85 was obtained with fine WO model powder. Therefore, a prospect to satisfy both desirable powder properties and simplified nuclear material production was opened, as well as improvement of working efficiency and cut down on costs.
Koma, Yoshikazu; Sano, Yuichi; Morita, Yasuji; Asakura, Toshihide
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.1056 - 1060, 2009/09
The chromatography technology has been investigated for actinides (An) (III) separation from a highly acidic solution as a part of reprocessing system for spent fast breeder reactor fuel. The adsorbent for the chromatography is based on the porous silica support impregnated with an extractant that is used for solvent extraction. Several extractants, CMPO, TODGA, HDEHP and BTP, were examined for the two steps flowsheet; An(III)/lanthanides (Ln) (III) recovery and An(III)/Ln(III) separation. The CMPO and TODGA adsorbents are the potential candidate from the chromatogram of Am, Cm and some fission product elements for An(III)-Ln(III) recovery whereas the HDEHP and BTP adsorbents are for An(III)/Ln(III) separation. The regenerated adsorbents after removing and re-impregnating the extractant show the identical adsorption property to the original.
-dodecane for U(VI) separationBan, Yasutoshi; Burdet, F.*; Cames, B.*; Caniffi, B.*; Hill, C.*; Morita, Yasuji
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.266 - 271, 2009/09
N,N-di(2-ethylhexyl)butanamide, N,N-di(2-ethylhexyl)isobutanamide, and N,N-di(2-ethylhexyl)dimethylpropanamide diluted to ca. 2 mol/dm
(M) in dodecane pre-equilibrated with 5 M HNO were degraded by 
-ray up to the integrated dose of ca. 1040 kGy. Identification of degradation products, and extraction of simulated fission products (Sr, Ba, Zr, Mo, Ru, Rh, Pd, and Nd) and U(VI) by the three degraded monoamides were respectively carried out. The three monoamides were practically stable against -ray irradiation, and these monoamides and their degraded products hardly have detrimental effects on the extraction separation of U(VI). Although the distribution ratios toward Pd increased with increasing integrated dose, they were less than unity up to the integrated dose of ca. 300 kGy. The three irradiated monoamides conserve their loading capacities of ca. 0.5 M as single stage extraction of U(VI), and the loading capacities decrease little with increasing integrated dose.
Morita, Yasuji; Kim, S.-Y.; Kawata, Yoshihisa; Ikeda, Yasuhisa*; Kikuchi, Toshiaki*
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.1081 - 1085, 2009/09
We have been developing an advanced reprocessing system for spent FBR fuels based on precipitation method using pyrrolidone derivatives, which consists of two precipitation steps; the first selective U precipitation step and the second U-Pu co-precipitation step. In the present study, precipitation behavior of Pu with pyrrolidone derivatives with high precipitation ability of N-(1,2-dimethyl)propyl-2-pyrrolidone (NDMProP), N-neopenthyl-2-pyrrolidone (NNpP) and N-cyclohexyl-2-pyrrolidone (NCP) has been examined with solutions of U-Pu mixture in order to evaluate their applicability to the second step. Since NNpP showed the highest precipitation ability for Pu(IV) and the best physical property as precipitate, NNpP would be the most appropriate precipitant for the U-Pu co-precipitation process. Precipitation of Np(IV, V, VI) with NNpP was also examined and it was found that Np(VI) could be quantitatively co-precipitated with U(VI) and Pu(IV).
Tezuka, Masashi; Mizui, Hiroyuki; Matsushima, Akira; Nakamura, Yasuyuki; Hayashi, Hirokazu; Sano, Kazuya; Nanko, Takashi; Morishita, Yoshitsugu
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.2815 - 2821, 2009/09
FUGEN is a proto-type heavy water moderated, boiling light water cooled, pressure tube type reactor with 165MWe and has been shut downed on Mar. 2003. Following the approval of decommissioning program in 2008, stage of FUGEN was changed to the decommissioning of the facilities. The program consists of following four periods; (1) Spent fuel transportation, (2) Periphery facilities dismantlement, (3) Reactor dismantlement and (4) Building demolition. It is expected that the whole decommissioning will be completed until 2028. As a part of the work in the spent fuel transportation period, the main steam system and the feeder water system etc. are being dismantled in the turbine building. The remaining tritium in the heavy water system is also being removed for facilitating the dismantlement of the heavy water system. Moreover, method on dismantlement of the reactor core is being studied with considering the process under the water for the radiation shielding and the dust suppression.
Uchiyama, Gunzo; Abe, Hitoshi
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.304 - 308, 2009/09
The simulation code, SAFE (Simulation code of anticipated transients formed extraction process), was developed in JAEA (Formerly JAERI) to obtain a highly functional tool based on a refined model that allows reliable safety assessment with reasonable safety margin of anticipated transient extraction behavior in reprocessing. Anticipated transient extraction behaviors of U and Pu have been studied with the SAFE code to understand the effect of process parameters in solvent extraction cycles in current large-scale reprocessing plants. In this paper the analytical results of transient extraction behaviors of Pu in the Pu purification cycle are mainly described. It was found that the concentration of Pu in the Pu stripping and Pu scrubbing steps increased with decreasing of the flow rate of nitric acid solution, which caused a Pu leakage from the Pu stripping to the Pu scrubbing and then the U stripping steps.