Osaka, Masahiko; Miwa, Shuhei; Tanaka, Kosuke; Akutsu, Yoko; Ikeda, Kaoru*; Mimura, Hitoshi*; Suzuki, Tatsuya*; Usuki, Toshiyuki; Yano, Toyohiko*
Annals of Nuclear Energy, 38(12), p.2661 - 2666, 2011/10
Novel concepts for effective utilization of molybdenum (Mo) from nuclear waste and magnesium silicates from hazardous asbestos wastes are proposed. A fast reactor cycle scheme that incorporates each material is described in the present paper. Basic studies on some fundamental technologies for the present cycle are given. Basic separation aspects for Mo by using LIX63 micro capsules and tertiary pyridine resin were investigated. A simple chemical synthesis route for Mo precursor powder from Mo containing HNO solution was tested. Effects of impurities in recovered Mo on sintering behavior were experimentally investigated.
Meiliza, Y.; Oki, Shigeo; Okubo, Tsutomu
Progress in Nuclear Energy, 53(7), p.964 - 968, 2011/09
Furukawa, Tomohiro; Inagaki, Yoshiyuki; Aritomi, Masanori*
Progress in Nuclear Energy, 53(7), p.1050 - 1055, 2011/09
Compatibility of the FBR candidate materials, 12Cr-steel and 316FR, with supercritical CO pressurized at 20 MPa for up to 8000 hours at 400-600 C has been investigated. Corrosion due to the high temperature oxidation was measured in both steels. Results showed that the behavior differed greatly. For 12Cr-steel, weight gain showed parabolic growth as exposure time increased at each temperature, and no breakaway oxidation was observed. The specimens were covered by two successive oxide layers. For 316FR, the weight gain was significantly lower than that of 12Cr-steel, and good resistance against corrosion was observed. No dependency of temperature or immersed time on weight gain was observed. Based on the metallurgical examination, corrosion formula in supercritical CO has been shown for the candidate materials for provisional design.
Usuki, Toshiyuki; Yoshida, Katsumi*; Yano, Toyohiko*; Miwa, Shuhei; Osaka, Masahiko
Progress in Nuclear Energy, 53(7), p.1078 - 1081, 2011/09
The effects of sintering additives of magnesium silicates, i.e. enstatite (MgSiO), steatite (MgSiO) and forsterite (MgSiO), on the fabrication properties and characteristics of the silicon nitride ceramics based inert matrix fuels were experimentally investigated. CeO was selected as simulating element of AmO. Sintered pellets were characterized in term of their densities, thermal conductivities and solubility to nitric acid. The densifications of sintered bodies were enhanced by using additives of magnesium silicates at relative low sintering temperature. The relative density of silicon nitride ceramics based inert matrix fuels with forsterite were achieved above 90% at 1723 K. The thermal conductivities of silicon nitride ceramics based inert matrix fuels varied according to sintering temperature, and those sintered at 1923 K were above 34 W/m K. The grain boundary phases in Silicon nitride ceramics based inert matrix fuels found to be dissolved into HNO.
Permana, S.; Suzuki, Mitsutoshi
Progress in Nuclear Energy, 53(7), p.958 - 963, 2011/09
Material attractiveness evaluation based on isotopic plutonium barrier compositions have been investigated based on intrinsic feature of proliferation resistance such as decay heat (DH), spontaneous fission neutron (SFN), as well as attractiveness concepts of figure of merit (FOM) and attractiveness concept (ATTR) as a function of diluted fraction of even mass plutonium to Pu-239 composition.
Ueta, Shohei; Aihara, Jun; Sawa, Kazuhiro; Yasuda, Atsushi*; Honda, Masaki*; Furihata, Noboru*
Progress in Nuclear Energy, 53(7), p.788 - 793, 2011/09
In Japan, high temperature gas-cooled reactor (HTGR) fuel fabrication technologies have been developed by Nuclear Fuel Industries, Ltd. (NFI) with the collaboration of JAEA through the HTTR project since 1960's. NFI successfully fabricated first and second loading fuel (0.9 tU each) for the HTTR of JAEA. Its excellent quality was confirmed from the first loading fuel through the long-termed high temperature operation by the end of March 2010. Based on the HTTR fuel technologies, silicon carbide (SiC) coated fuel is being developed for burn-up extension. For an advanced fuel designs, replacement of the SiC layer by a zirconium carbide (ZrC) layer is a very promising example. JAEA has performed ZrC coating tests to investigate the influence of coating parameters and material properties such as stoichiometry and density of ZrC.
Miwa, Shuhei; Osaka, Masahiko; Usuki, Toshiyuki; Yano, Toyohiko*
Progress in Nuclear Energy, 53(7), p.1045 - 1049, 2011/09
We proposed a new concept for densification of minor actinides-containing inert matrix fuels by using asbestos waste-derived materials for the effective utilization of resources and protection of public safety. In this concept, magnesium silicates, which are mainly generated by the decomposition of asbestos in low temperature heat-treatment, are used as a sintering additive for the achievement of high density of magnesia-based inert matrix fuels at relatively low sintering temperature. In this study, preliminary fabrication tests of magnesia-based inert matrix fuels with magnesium silicates were carried out by using cerium oxides as a representative of minor actinides oxides.
Koyama, Shinichi; Suzuki, Tatsuya*; Mimura, Hitoshi*; Fujita, Reiko*; Kurosawa, Kiyoko*; Okada, Ken*; Ozawa, Masaki
Progress in Nuclear Energy, 53(7), p.980 - 987, 2011/09
Individual basic researches of separation step were performed in the Advanced ORIENT Cycle project. High separation selectivity for Cs and Sr by novel nano adsorbents AMP-SG (D) and D18C6-MC were confirmed, respectively. TPR well adsorbed Pd and Tc in dilute HCl condition. Formation of rare metal fission product RMFP-deposit Pt electrodes from SHLLW was verified, and it was confirmed that high catalytic reactivity on electrolytic production of hydrogen. As experiment for engineering feasibility, Hastelloy-B at RT and Ta at 90C were confirmed their anti-corrosive in highly concentrated HCl media. Thermo-chemical stability for TPR was verified in either HCl or HNO media toward its practical use in the separation process. Issues to be solved for optimization based on the results of lab-scale experiment have revealed in this study.
Nogami, Masanobu*; Harada, Masayuki*; Sugiyama, Yuichi*; Kawasaki, Takeshi*; Kawata, Yoshihisa*; Morita, Yasuji; Kikuchi, Toshiaki*; Ikeda, Yasuhisa*
Progress in Nuclear Energy, 53(7), p.948 - 951, 2011/09
The precipitation ability of 1,3-dimethyl-2-imidazolidone (DMI) to U(VI) and U(IV) was examined using nitric acid solutions. While DMI precipitated U(VI) from 3 M nitric acid, no precipitate was observed in the solution containing 0.15 M U(IV) and 3 M nitric acid by adding DMI at the ratio of [DMI]/[U(IV)]=5. This indicates that the selectivity of DMI to U(VI) than U(IV). In spite of the excellent selectivity to U(VI), DMI has a disadvantage on the stability in nitric acid, because gradual acid hydrolysis of DMI is inevitable due to the nature of the chemical structure. Experiments on the stability of DMI in -ray irradiation and heating in nitric acid solutions showed that the stability is strongly affected by the concentration of nitric acid and that DMI may be applicable in nitric acid solutions up to ca. 2 M.
Obara, Toru*; Yamazawa, Yu*; Sasa, Toshinobu
Progress in Nuclear Energy, 53(7), p.1056 - 1060, 2011/09
Lead-Bismuth eutectic (LBE) has many good characteristics as a coolant for fast reactors. One of the issues remeining to be solved, however, is the polonium issue. The purpose of the present study was to estimate the decontamination performance of a polonium filter by experiment in the penetration condition. Two types of stainless steel wire meshes, fine wire mesh and loose wire mesh, were tested in the experiments. The results show that polonium filters made of stainless steel wire mesh can be very useful device for the removal of polonium in the gas phase. These filters can be used for the decontamination of primary loops by the baking method.
Proceedings of 3rd International Symposium on Innovative Nuclear Energy Systems (INES-3), 10 Pages, 2010/11
(1) The cost of electricity generated by 54 Japanese light water reactors in 2005 is 42,682 million dollars (2) During a nuclear cycle the emitted carbons from LWR (22 g/kWh) is 1/23 to 1/44 of those from fossil power plants. The gross electricity produced in Japan in 2004 is about 8,651 TWh. Emitted carbons assuming that coal and petroleum are main carbon contributors are 7.43E08 ton carbon dioxide. The non-fossil fuels can suppress the amounts by 3.79E08 ton carbon dioxide, where the contributing ratio of nuclear energy is 57%. Because the price of carbon emission trading is 18.5 dollars for the developing country and 27.7 dollars for European Community (EC) per ton carbon dioxide, an indirect effect of green technology by Japanese LWRs is estimated to be 3,993 million dollars for the former and 5,989 million dollars for the latter. Consequently, the role of green technology to suppress the global warming is very significant and cost effective.
no journal, ,
Without distinction of expertise such as reactor physics or thermal-hydraulics, mathematical models explaining physical phenomena use differential equations and have been changed from homogenous to heterogeneous. However, considering time cost until obtaining desire analytical results, homogeneous models have enough convenience today. Therefore, differential equations explaining physical phenomena include homogeneous and heterogeneous variables. In this paper, methodology for quantitative evaluation by using sum of factorizations is presented. This method is useful to cognize physically the analytical difference between homogeneous model and heterogeneous one because general method can show only numerical value without relationship among distributions of variables. Regardless of reactor physics or thermal-hydraulics, differential equations can be comprehended from a new different viewpoint.
Sato, Hiroyuki; Kubo, Shinji; Yan, X.; Tachibana, Yukio; Kato, Yukitaka*
no journal, ,
One of the key areas to be investigated in VHTR co-generation system is an operability during transients initiated by the abnormal events by the hydrogen production plant since a continuous reactor operation and electricity generation is required in case of abnormal events in the hydrogen production plant from the economical point of view. In this study, control strategies are proposed and investigated against the transients initiated in the hydrogen production plant. Various transients were evaluated using the newly developed dynamic simulation code based on the RELAP5. The simulation results showed that the reactor could operate continuously during the transients of hydrogen production plant employing the proposed control strategy. In addition, the advantages of combining several control strategies are discussed comparing to the single application.
Ishiyama, Koichi; Yamamura, Osamu
no journal, ,
Tokai Reprocessing Plant (TRP) started the hot operation test in 1975 and had safely and stably reprocessed for more than 30 years, and completed the reprocessing of 1116 tons of spent fuels (SF) by contract for service with electric power companies in the spring of 2006. During the reprocessing operations, TRP had made every effort to validate the nuclear proliferation resistance technique for plutonium, etc.. First of all, as part of technical development for nuclear material accountancy (MA) and SG, TRP developed Solution Measurement and Monitoring System to confirm no abnormal proliferation of nuclear material. Moreover, IAEA appreciated the transparent MA by TRP, because TRP has developed many SG technologies and cooperated in a positive manner to make a success of various IAEA's inspections through the past reprocessing operations. As a result, Integrated Safeguards has been introduced into TRP since 2008 and TRP is the first reprocessing facility worldwide to have attained it.
Onishi, Takashi; Obayashi, Hiroshi; Koyama, Shinichi; Mimura, Hitoshi*
no journal, ,
Hybrid microcapsule (MCs) is composed of carrier matrix and extractants and form granulated composite, so that it is suitable for the practical column operation. Previous studies revealed the uptake behaviors of various metal ions on MCs. In this study, the selective uptake and recovery from simulated high level radioactive liquid waste solution (HLLW) were examined. Firstly, the uptake experiments of Cs, Pd, Re substitute for Tc and Mo were carried out by batch method using seven kinds of MCs. As a result, the uptake rates for all MCs were more than 96.9 %. Secondly, the column separation experiments were conducted with four MCs for the separation of Cs, Pd, Re and Mo, respectively. The results indicate the strong affinity of Cs, Pd and Re on each microcapsule, and it is demonstrated that the selective uptake and recovery from simulated HLLW by column method with MCs, and their availability was revealed for the treatment of HLLW.
Imai, Yoshiyuki; Wang, L.*; Guo, H.*; Kasahara, Seiji; Kubo, Shinji
no journal, ,
The purification process of HIx phase was analysed using process simulator ESP and thermodynamic data OLI-MSE model. Desulfurization ratio increased largely with increasing stripping gas flow rate. Heat duty depended little on stripping gas flow rate, but, increased largely with temperature. Purification can be carried out with low heat duty at the same temperature as that of HIx by using stripping gas.
Terada, Atsuhiko; Shimakawa, Satoshi; Shibata, Taiju; Shiozawa, Shusaku*; Minatsuki, Isao
no journal, ,
High Temperature Gas-cooled Reactor (HTGR) has several features different from conventional light water reactors such as high passive safety characteristics, high thermal efficiency and high economy for small sized reactor with modular concept. On the other hand, one of disadvantages of HTGR with prismatic core is to require rather long-term and expensive refueling, resulting in relatively high maintenance period and cost. To solve the disadvantage, the present study challenges the core design of MHR-50 for long refueling interval by increasing core size, fuel loading and fuel burn-up. The preliminary burn-up calculation suggested that approximately 10 years of long refueling interval was found to be reasonably achieved.