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irradiated with high-energy heavy ionsIshikawa, Norito; Sonoda, Takeshi*; Okamoto, Yoshihiro; Sawabe, Takashi*; Takegahara, Keisuke; Kosugi, Shinya*; Iwase, Akihiro*
Journal of Nuclear Materials, 419(1-3), p.392 - 396, 2011/12
Times Cited Count：11 Percentile：62.72(Materials Science, Multidisciplinary)In order to characterize the radiation damage due to ion-track formation in UO, polycrystalline samples have been irradiated with 210-MeV Xe ions, and measured with XRD (X-ray diffraction) technique using Cu X-ray. We have also tried EXAFS (extended X-ray absorption fine structure) measurement using X-ray near U L
-edge. The results show that XRD technique detects damage at relatively low fluence of 10
ions/m
and higher, while the irradiation-induced change of EXAFS spectra is not observed even at highest fluence of 10
ions/m. The damage detection may be critically influenced by the depth profile of X-ray penetration.
Yano, Yasuhide; Ogawa, Ryuichiro; Yamashita, Shinichiro; Otsuka, Satoshi; Kaito, Takeji; Akasaka, Naoaki; Inoue, Masaki; Yoshitake, Tsunemitsu; Tanaka, Kenya
Journal of Nuclear Materials, 419(1-3), p.305 - 309, 2011/12
Times Cited Count：22 Percentile：81.76(Materials Science, Multidisciplinary)The effects of fast neutron irradiation on ring tensile properties of oxide dispersion strengthened (ODS) steel claddings for fast reactor were investigated. Specimens were irradiated in the experimental fast reactor Joyo using the material irradiation rig at temperatures between 693 and 1108 K to fast neutron doses ranging from 16 to 33 dpa. The post-irradiation ring tensile tests were carried out at irradiation temperatures. The experimental results showed that there was no significant change in tensile strengths after neutron irradiation below 923 K, but the tensile strengths at neutron irradiation above 1023 K up to 33 dpa were decreased by about 20%. On the other hand, uniform elongation after irradiation was more than 2% at all irradiation conditions. The ring tensile properties of these ODS claddings remained excellent within these irradiation conditions compared with conventional 11Cr ferritic/martensitic steel (PNC-FMS) claddings.
Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi
Journal of Nuclear Materials, 414(2), p.303 - 308, 2011/07
Times Cited Count：12 Percentile：65.69(Materials Science, Multidisciplinary)In order to estimate the behavior of high burnup mixed-oxide (MOX) fuel, it is important to evaluate fuel temperature accurately. The thermal conductivity formula of MOX fuel pellet which is needed to evaluate the fuel temperature was proposed. By using Klemens's theory and reported thermal conductivities of unirradiated (U, Pu)O and irradiated UO pellets, the thermal conductivity formula which contains the effects of burnup and plutonium (Pu) addition was obtained. Temperature of high burnup MOX fuel was evaluated based on the above-mentioned formula and the thermal conductivity integral method, and was compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it is considered that the proposed thermal conductivity formula of MOX pellets is adequate.
,Pu
)O
Takeuchi, Kentaro; Kato, Masato; Sunaoshi, Takeo*
Journal of Nuclear Materials, 414(2), p.156 - 160, 2011/07
Times Cited Count：11 Percentile：62.72(Materials Science, Multidisciplinary)The sintering behavior of MOX pellets was investigated by thermal gravimetry and thermal dilatometry. The starting material was prepared by the microwave direct heating de-nitration method, in which the Pu/(Pu+U) ratio was controlled to 20% in the nitrate solution. The powder was pressed into sample pellets by the die wall lubrication method. The two kind of test, the constant heating rate test and the isochronal heating test, were carried out in various sintering atmospheres of 
/
ratio. The results of the constant heating rate test showed that the shrinkage rate and O/M ratio increased with decreasing the / ratio. The isochronal heating test was carried out in the O/M range of 1.98 - 2.0005, and the densification behavior of the pellets was analyzed by use of the equation; y = (
/
) = 
(
)
. The result showed that the sintering mechanisms varied with the O/M ratio and temperature.
Pu
)O and (U
Pu
)O at high temperatures of 1673-1873 KKato, Masato; Takeuchi, Kentaro; Uchida, Teppei; Sunaoshi, Takeo*; Konashi, Kenji*
Journal of Nuclear Materials, 414(2), p.120 - 125, 2011/07
Times Cited Count：22 Percentile：82.80(Materials Science, Multidisciplinary)Many studies on oxygen potentials have been reported, but their data were scattered and the data at high temperatures are limited. In this work, the oxygen potential of (UPu)O and (UPu)O was measured at high temperatures of 1673-1873 K using gas equilibrium method using thermo-gravimetry. The influence of Pu addition on the oxygen potential of MOX was discussed. The oxygen potential and the O/M ratio were decided by in-situ analysis. The oxygen partial pressure was adjusted by controlling the ratio of 
/O in the flowing gas atmosphere, and the oxygen potential was determined. The oxygen potentials measured by the point defect model. The deviation x varied with the relation of in the near stoichiometric composition region. The oxygen potential increased with increasing Pu content. The values of stoichiometric MOX containing 12% and 30%Pu were determined to be -334 kJ/mol and -296 kJ/mol, respectively, at 1773 K.
at temperatures from 300 to 1500 KTanaka, Kosuke; Sato, Isamu; Hirosawa, Takashi; Kurosaki, Ken*; Muta, Hiroaki*; Yamanaka, Shinsuke*
Journal of Nuclear Materials, 414(2), p.316 - 319, 2011/07
Times Cited Count：24 Percentile：84.88(Materials Science, Multidisciplinary)Polycrystalline specimens of barium plutonate, BaPuO, have been prepared by mixing the appropriate amounts of PuO and BaCO followed by reacting and sintering at 1600 K under the flowing gas atmosphere of dry-air. The thermal conductivity of BaPuO was almost the same magnitude as that of BaUO.
Nagai, Takayuki; Uehara, Akihiro*; Fujii, Toshiyuki*; Sato, Nobuaki*; Yamana, Hajimu*
Journal of Nuclear Materials, 414(2), p.226 - 231, 2011/07
Times Cited Count：11 Percentile：62.72(Materials Science, Multidisciplinary)Nagae, Yuji; Takaya, Shigeru; Wakai, Eiichi; Aoto, Kazumi
Journal of Nuclear Materials, 414(2), p.205 - 210, 2011/07
Times Cited Count：3 Percentile：25.34(Materials Science, Multidisciplinary)Murakami, Tsuyoshi*; Sakamura, Yoshiharu*; Akiyama, Naoyuki*; Kitawaki, Shinichi; Nakayoshi, Akira; Fukushima, Mineo
Journal of Nuclear Materials, 414(2), p.194 - 199, 2011/07
Times Cited Count：17 Percentile：76.53(Materials Science, Multidisciplinary)An electrorefining is one of the main steps of pyrochemical reprocessing of spent metallic fuels (U-Zr, U-Pu-Zr). The electrorefining is carried out dissolving a portion of Zr together with actinides to accomplish a high dissolution ratio of actinides. However, the electrorefining with Zr co-dissolution should bring some practical problems in the pyrochemical reprocessing. Therefore, electrorefining tests of non-irradiated U-Pu-Zr alloy were performed with minimizing the amount of Zr dissolved in LiCl-KCl-(U, Pu, Am)Cl melts at 773 K. The tests were performed both by potentiostatic electrolysis at -1.0 V (Ag
/Ag) that was more negative than the Zr dissolution potential and by galvanostatic electrolysis with a limited amount of Zr dissolution. The ICP-AES analysis of the anode residues confirmed that a high dissolution ratio of actinides (U; 
99.6%, Pu; 99.9%) was successfully demonstrated at both electrolyses.
Sekio, Yoshihiro; Yamashita, Shinichiro; Yoshitake, Tsunemitsu; Nishinoiri, Kenji; Takahashi, Heishichiro
no journal, ,
The process of the irradiation-induced defects accumulation during irradiation has not been clarified yet, namely what microstructural factors affect defect accumulation process. For this reason, we investigated the void distribution near grain boundaries by focusing on grain boundaries as a microstructural factor in Fe-15Cr-15Ni model alloy and PNC316 stainless steel used for the nuclear fast reactor. It was clarified that the void denuded zone were formed near only a random grain boundary under neutron irradiation in Fe-15Cr-15N alloy and PNC316 stainless steel as being observed in the electron-irradiated austenitic stainless steels. We discussed the difference of the void denuded zone distribution in both steels under low and high fluencie. These results suggest that the grain boundary characteristics and its effect on void denuded zone formation will play a role for the void swelling under neutron irradiation in austenitic stainless steels.
Miwa, Shuhei; Osaka, Masahiko
no journal, ,
Reduction-oxidation behaviors and oxygen potentials of plutonium oxide in magnesia-based inert matrix fuels were experimentally investigated by thermogravimetry. The reduction and oxidation rates of plutonium oxide in magnesia-based inert matrix fuels were lower than those of plutonium oxide. The oxygen potentials of plutonium oxide-containing magnesia-based inert matrix fuels were lower than those of plutonium oxide near stoichiometry, and became closer to those of plutonium oxide with decreasing oxygen-to-metal ratio. These results indicated that the thermodynamic and static properties of inert matrix fuels could be changed for the better by using the MgO matrix.
under accumulation of ion tracks by high energy fission productsSonoda, Takeshi*; Ishikawa, Norito; Sataka, Masao; Sawabe, Takashi*; Kitajima, Shoichi*; Kinoshita, Motoyasu*
no journal, ,
Microstructural evolutions of ion-irradiated UO were observed by use of a 300 kV FE-TEM and a FE-SEM at CRIEPI. For understanding the effects of accumulation of ion tracks in UO, 210 MeV Xe+16 ions irradiation up to a fluence of 1.5 e16 ions/cm were done. At a fluence of 5e14 ions/cm (overlapping num. of tracks: 1e2), elliptical changes of fabricated pores and the formation of dislocations are started. Over 1e16 ions/cm (overlapping num. of tracks: 2e3), the grain sub-divisions are observed and the dislocation density does not tend to increase. These drastic changes in UO indicate that the overlapping of ion tracks will cause the point defects, enhance the diffusion of point defects and dislocations, and form the sub-grains at relatively low temperature.
Tachi, Yoshiaki; Wakabayashi, Toshio*
no journal, ,
For reducing the environmental burden and toxic risk of high-level radioactive wastes, it is one of the effective techniques to transmute iodine-129 by fast reactors. Since iodine has corrosive properties, the chemical form of the iodine-bearing transmutation target material loaded into the FR is one of the important issues. In this study, from the viewpoints of melting point and neutron irradiation properties, the following five iodides were selected as the candidate materials from all binary iodides: CuI, MgI, YI, RbI and BaI. In order to clarify their applicability as a transmutation target, TG-DTA (Thermogravimetry - Differential Thermal Analysis) of the iodides and compatibility of the iodides with the stainless-steel cladding materials were tested. From the results of them, BaI is preferable as a transmutation target material.
Yamashita, Shinichiro; Yano, Yasuhide; Akasaka, Naoaki; Otsuka, Satoshi; Kaito, Takeji; Inoue, Masaki
no journal, ,
Neutron irradiation behavior of 9Cr- and 12Cr-ODS steels cladding tubes developed for fast reactor was investigated to understand the effect of neutron irradiation on their microstructures. These ODS steels cladding tubes were irradiated at fuel-free condition and at 683 to around 1,100 K to fast neutron fluences ranging from 3.0 to 6.6 
10
n/m (E0.1 MeV) in JOYO. Metallurgic examination and TEM observations revealed that the radiation-induced defect cluster formation was suppressed and that each matrix structure of 9Cr- and 12Cr-ODS steels possessed relatively high microstructural stability during irradiation at elevated temperature. These results attributed to a high density of defect sink site existed in each of their initial matrices and also to microstructural stabilization effect of thermally stable oxide dispersion, that is, the suppression effects of oxide on microstructural evolutions, such as dislocation recovery and grain growth, during irradiation.
Aoyagi, Yoshiteru; Igarashi, Takahiro; Kaji, Yoshiyuki
no journal, ,
Stress corrosion cracking (SCC) is one of the critical concerns as degradation of structural components in light water reactors (LWRs) for long period. Many studies on intergranular SCC (IGSCC) have been conducted over several decades, however the mechanism of IGSCC initiation and propagation is not fully understood. In this study, in order to investigate the mechanism of IGSCC, we propose the multi-physics model with the combination of oxygen atom diffusion model and crystal plasticity model for IGSCC. Using this model, we conduct the simulation of crack propagation for polycrystalline metals. The conclusions obtained here are summarized as follows: (1) In the present model, it is possible to reproduce that IGSCC cracks propagate with branching by repeating growth and stop processes of micro cracks. (2) The behavior of oxygen along the grain boundary might be one of the important factors which determine the geometry of IGSCC propagation.
Minato, Kazuo; Konashi, Kenji*; Fujii, Toshiyuki*; Uehara, Akihiro*; Nagasaki, Shinya*; Otori, Norikazu*; Akabori, Mitsuo; Takano, Masahide; Hayashi, Hirokazu; Tokunaga, Yo; et al.
no journal, ,
Basic research in actinide chemistry and physics is indispensable to maintain sustainable development of innovative nuclear technology. Actinides, especially minor actinides (MA) of americium and curium, need to be handled in special facilities with containment and radiation shields. A three-year-program Basic actinide chemistry and physics research in close cooperation with hot laboratories, ACTILAB, was started to form the basis of sustainable development of innovative nuclear technology. For the basic understanding of the properties of MA-bearing fuels, XAFS (X-ray Absorption Fine Structure) measurements on AmO
and 
O-NMR (Nuclear Magnetic Resonance) measurements on (Pu
Am
)O were carried out. Curium nitride was synthesized by carbothermic reduction method and the lattice parameters and thermal expansion of CmN were measured by high temperature X-ray diffractometry after purification of aged curium oxide.
molten saltKurata, Masaki*; Murakami, Tsuyoshi*; Kitawaki, Shinichi; Nakayoshi, Akira; Fukushima, Mineo
no journal, ,
CRIEPI and JAEA have continued a collaboration study for pyro-reprocessing. Various kinds of sequential electrolysis are performed in an 1.2 kg of molten salt bath that includes U, Pu and Am. Recent results are reported, in which several batches of U-Zr alloy anode and liquid-Cd or solid-Fe cathode were continuously used. Variation in current efficiency, compositions of cathode product, anode residue and molten salt, variation in mass balance, etc. were measured successively based on two kinds of the electrolysis procedure.
Ebihara, Kenichi; Yamaguchi, Masatake; Nishiyama, Yutaka; Onizawa, Kunio; Matsuzawa, Hiroshi*
no journal, ,
The experimental results on neutron-irradiated reactor pressure vessel (RPV) steels have revealed grain boundary (GB) segregation of phosphorous(P) causing GB embrittlement. Since the dependence of the segregation on dose and dose-rate is unclear due to the lack of experimental database, the estimation of the irradiation-induced GB P segregation by the rate theory model is desired. In this presentation, we incorporated the effect of carbon(C) into the model by considering C as the trap of vacancies(Vs) and self-interstitial atoms (SIAs), and evaluated the GB P coverage in the RPV steels. As a result, by selecting the proper sink strength of V and SIA, the simulation reproduced the experimental GB P coverage. It was confirmed that C hardly influences the dose rate dependency of the GB P coverage. It was found that C influences the GB P coverage by mainly slowing the V migration. Therefore it is considered that V is necessary for simulating GB P coverage even though V hardly transports P.
Suzuki, Chikashi; Nishi, Tsuyoshi; Nakada, Masami; Tsuru, Tomohito; Akabori, Mitsuo; Hirata, Masaru; Kaji, Yoshiyuki
no journal, ,
no abstracts in English
lattice and bulk expanded by self-irradiation damageTakano, Masahide; Akabori, Mitsuo; Arai, Yasuo
no journal, ,
To investigate the influence of self-irradiation damage and accumulation of He in oxide fuel pellets containing minor actinides, the expansion and annealing behavior of the crystal lattice and bulk were examined comparatively on a (Pu
Cm
)O specimen. The lattice parameter and pellet dimension at room temperature grew with the similar dependence on storage time, and their linear expansion saturated to 0.3 % within 60 days. After two years storage, the annealing examinations were performed in the temperature range up to 1433 K. The lattice parameter recovered to the undamaged value, whereas the pellet dimension grew again during the heat treatment at 1433 K. The fractured surface of the annealed pellet was then observed by a scanning electron microscope. The gas bubbles with 200-300 nm in diameter were recognized along the grain boundaries like the fission gas bubbles. It was considered the dissolved He atoms moved toward grain boundaries by diffusion and formed gas bubbles, which resulted in the He gas swelling of the pellet specimen.
