Osaka, Masahiko; Donomae, Takako; Ichikawa, Shoichi; Sasaki, Shinji; Ishimi, Akihiro; Inoue, Toshihiko; Sekio, Yoshihiro; Miwa, Shuhei; Onishi, Takashi; Asaka, Takeo; et al.
Proceedings of 1st Asian Nuclear Fuel Conference (ANFC), 2 Pages, 2012/03
Support system for training and education of future expert in hot laboratories of Oarai-JAEA, named FEETS, is presented. The system has been established based on research results on both characterization of Oarai hot laboratory and user-needs. Various programs under FEETS are also introduced.
Donomae, Takako; Ito, Masahiko*
Journal of Nuclear Science and Technology, 48(5), p.826 - 833, 2011/05
The release and retention of tritium generated in FBR control rods of BC irradiated to high fluence were investigated for temperatures up to 800C. The amounts of tritium retention were expressed as a function of fluence, and the results agreed approximately with the theoretically calculated values. In addition, the amounts of tritium could be approximately evaluated by simply measuring the radioactivity of Mn using germanium semiconductor detector, which is generated from Fe impurity in BC. The temperature at which the tritium begins to be released from BC was investigated by heating a specimen at a constant temperature increasing rate. It was found in this experiment that the tritium begins to be released from BC at 400C. From the results of this isochronal heating, diffusion coefficients of tritium in BC were derived.
Donomae, Takako; Katsuyama, Kozo; Tachi, Yoshiaki; Maeda, Koji; Yamamoto, Masaya; Soga, Tomonori
Journal of Nuclear Science and Technology, 48(4), p.580 - 584, 2011/04
One of the challenges in developing a long-life control rod is to restrain absorber-cladding mechanical interaction (ACMI). Its lifetime was limited by ACMI, which is induced by the swelling and relocation of BC pellets. To restrain ACMI, a shroud tube was inserted into the gap between the BC pellets and the cladding tube. And sodium was selected as bonding material instead of helium to restrain increases in the pellet temperature. As a result of these improvements, the estimated lifetime of the control rod at Joyo was doubled. In this paper, the results of post irradiation examination are reported.
Uwaba, Tomoyuki; Sogame, Motomu; Ito, Masahiro*; Mizuno, Tomoyasu; Donomae, Takako; Katsuyama, Kozo
Journal of Nuclear Science and Technology, 47(8), p.712 - 720, 2010/08
In determining lifetime criteria of fast reactor fuel pins, creep damage due to fission gas pressure on mixed-oxide fuel pins with austenitic stainless steel cladding successfully irradiated to high burnups (120 GWd/t or higher pin averaged burnup) was evaluated. The degree of creep damage of these fuel pins was expressed as cumulative damage fractions (CDFs), defined so that cladding breaching occurs when the CDF exceeds 1.0. The obtained CDFs for typical high temperature fuel pins were on the order of 10-10 at the end of irradiation, indicating that these fuel pins had large safety margins against breaching due to creep damage. In order to investigate the factors that govern the lifetime of fuel pins, pin diametral increase as well as CDF were predicted in cases of extended burnups from 120 GWd/t onward, and then were compared with tentatively determined limit values. The predicted pin diametral increase reached its limit value earlier than the CDF because of a significant increase in the cladding void swelling, suggesting that lifetimes of fuel pins with austenitic stainless steel cladding are practically governed by the diametral increase rather than by the creep damage.
Morohashi, Yuko; Maruyama, Tadashi*; Donomae, Takako; Tachi, Yoshiaki; Onose, Shoji
Journal of Nuclear Science and Technology, 45(9), p.867 - 872, 2008/09
Donomae, Takako; Tachi, Yoshiaki; Sekine, Manabu*; Morohashi, Yuko; Akasaka, Naoaki; Onose, Shoji
Journal of the Ceramic Society of Japan, 115(1345), p.551 - 555, 2007/09
Use of moderator materials in Fast Breeder Reactor (FBR) is effective for transmutation technology, and BC is one of the candidates. Up to now, the behavior of BC as the Control rod material is well known, but that of BC is hardly investigated. In this paper, the radiation effects of BC pellets, neutron irradiated in the experimental fast reactor JOYO were studied. From the experimental results, it was observed that no macro-cracks were recognized in the irradiated BC pellets. But, bubble nucleation was found in grain and along grain boundaries of BC. And, it was shown that the conductivity of BC was higher than that of BC. During the annealing from room temperature to 1400C, three recovery stages were found on thermal conductivity. It was suggested that, the recovery of BC was related to the dispersion behavior of helium. Judging from these results, as BC was mechanically more stable compared with BC under irradiation, it was shown that BC had high applicability for a moderator.
Donomae, Takako; Tachi, Yoshiaki; Matsumoto, Shinichiro
JAEA-Research 2006-033, 35 Pages, 2006/07
no abstracts in English
Osaka, Masahiko; Yoshimochi, Hiroshi; Donomae, Takako; Inoue, Masaki
JNC TN9420 2003-002, 42 Pages, 2003/10
Selection study of Am target candidates for the irradiation in the experimental fast reactor JOYO was performed by the investigation of previously reported studies as a part of JNC-JAERI joint project. Criteria of the selection of inert matrix for the Am target was decided and preliminary selection was carried out on the basis of the investigation results of the present status of development of target and several selection studies in the world. Detailed investigation of the properties such as thermal conductivity, irradiation behavior,fabrication technique on seven pre-selected candidates, specifically ZrO2, Al2O3, MgO, SiC, Si3N4, MgAl2O4 and TiN, was then performed. Five matrices for Am target candidates, ZrO2, MgO, SiC, Si3N4 and TiN,were consequently selected on considering the criteria,in particular, adaptability of the fabrication process were also designed and decided as follows: Dispersed type of Am oxide particle (dia:100/300 micro meter)into MgO (1) Am oxideparticle fabrication by dry route ; i.e. compacting of Am oxide powder, crushing and sieving (2) Mixing of Am oxide particle with MgO particle made by spray dry (3) Pelletizing by uni-axial press and sintering in one atm of inert atmosphere Mixed type of Am oxide (1) Mixing of Am oxide powder with material (SiC, Si3N4 or TiN) powder (2) Pelletizing by uni-axial press and sintering in one atm of inert atmosphere Fabrication test and property measurement will be conducted for the final decision of Am-containing target.
Donomae, Takako; ; Onose, Shoji; Miyakawa, Shunichi; Nakamura, Yasuo
JNC TN9420 2002-003, 19 Pages, 2002/03
The Partitioning and Transmutation (P&T) for radionuclides included in high level has been researched in many countries. This technology for the radionuclides consists of partitioning them to several groups according to their half-lives and purposes of utilization and transmutating minor actinides (MA) and long lived fission products (LLFP) to short lived or stable nuclides. Japan Nuclear Cycle Development Institute (JNC) made a plan to develop this technology in the Feasibility Study for Fast Reactors and Related Fuel Cycle (FS), in cooperation with basic research groups. The main objective of JNC is to transmutate MA and LLFP in fast reactor. And this research was planned to carry out, taking into account not only reduction of environmnental burden and nuclear non-proliferation but also technical realization and economics. As a part of the research, the development of the elements for irradiation tests has just stated. According to the gained results of FS. The LLFPs, which have a possibility to realize the transmutation from the view point of nuclear physics, are I and Tc. Therefore, it was tried to select iodine chemical compounds fitted for transmutation by means of literature survey, because the half-life of I is long and the effect of radiation is comparatively hard. The literature survey was performed from the viewpoint of five properties, that is, nuclear physics, thermal phase change, chemical stability, fabrication, applicability to cycle use. As a result, 8 chemical compounds, namely, MgI, KI, NiI, CuI, RbI, YI, MoI, BaI were selected as target materials from 32 candidates.
Tsai, H.*; Allen, T. R.*; Cole, J. I.*; Strain, R. V.*; Yoshitake, Tsunemitsu; Donomae, Takako; Akasaka, Naoaki; Mizuta, Shunji; Ukai, Shigeharu; Miyakawa, Shunichi
JNC TY9400 2001-025, 117 Pages, 2001/07
Donomae, Takako; Akasaka, Naoaki; Yamagata, Ichiro
JNC TN9400 2001-092, 44 Pages, 2001/03
; Donomae, Takako; *
Heavy irradiation ef, 0 Pages, 2001/00
Yoshitake, Tsunemitsu; Donomae, Takako; Mizuta, Shunji; James J.Co*
Progtams and Abstrac, p.469 - 486, 2001/00
Koyama, Akira*; Donomae, Takako; Kato, Yutai
JNC TY9400 2000-017, 65 Pages, 2000/03
no abstracts in English
; Yamagata, Ichiro; Donomae, Takako; Akasaka, Naoaki
JNC TN9400 2000-046, 24 Pages, 2000/02
lt is well known that solute atoms are segregated on surface, grain boundary, etc. and composition changed partially in irradiated austenitic stainless steel. For understanding radiation induced segregation (RIS), we adopt a Fe-15Cr-20Ni-x (x: Si, Mo) which is basically alloy system in PNC1520, and size of Si, Mo are different from matrix atoms to investigate RIS behaviors. The specimens were irradiated by "Joyo" fast reactor that irradiation condition is 3.5 10 n/m (E>0.1Mev) at 476C. After irradiation, the specimen were observed and analyzed with EDS (Energy Dispersive X-ray Spectroscope) of 400kV TEM (Transmission Electron Microscope). The behavior of RIS depends on size of solute atoms of alloy. For example, oversized atoms are decreased and undersized atoms are increased in sink. RIS of voids are as same as or more than grain boundaries and smaller than precipitates. The void denuded zone was existed nearby G.B. in case of combinations between the grains from G.B.0ne of the reasons in this, the voids swepted by moving G.B. in radiation induced G.B. migration.
Donomae, Takako; ; ; Akasaka, Naoaki; Yamagata, Ichiro; ;
JNC TN9400 2000-075, 374 Pages, 1999/08
The irradiation test of the Monju-type fuel subassembly, MEA-1, was conducted at FFTF as a Joint Research Program of Fuels and Materials between DOE and PNC. MFA-1 subassembly is consisted of cladding tube, wrapper tube, wrapping wire which are all made of PNC316, as well as 85% low density pellet. The pellet peak burnup and maximum fast neutron fluence reached 147.1GWd/t and 21.410n/m, respectively. Based on the results of Post-irradiation examination, Subassembly and fuel elements behaviors have been evaluated, and the following results were obtained. (1)The wrapper tube elongation and dilation are relatively small, and the fuel pin diameter changes were measured to be about 4% in maximum. The Bundle-Duct lnteraction (BDI) was confirmed not to be severe condition in Monju-type fuel assembly up to the fast neutron fluence of 21.410n/m. (2)In the claddings with the sand-brusted treatment to shave the flaw on the inner surface, the enhanced sweIling was measured, compared with PNC316 claddings manufactured by the usual process. lt is considered that this swelling enhancement in sand-brusted claddings attributed to relatively higher level of residual stress and lower cold-worked. (3)The deferent temperature dependency in swelling was evaluated for cladding and wrapping wire: the peak swelling temperature to be 495C for cladding and 475C for wrapping wire. 0n the contrary, the swelling-temperature dependency for wrapper tube was not able to be determined.
Koyama, Akira*; Donomae, Takako
JNC TY9400 2000-002, 286 Pages, 1999/03
no abstracts in English
; Akasaka, Naoaki; Mitsugi, Takeshi; Donomae, Takako; Ukai, Shigeharu;
ANS Winter Meeting, ,
Tachi, Yoshiaki; Donomae, Takako; Akasaka, Naoaki
no journal, ,
In order to transmutate of radioactive iodine, which is one of long lived fisson products, by using a fast reactor, we have investigated the most suitable chamical composition of iodine for loading in a reactor. We have conducted compatibility experiment to evaluate interaction between five iodides and cladding materials at 600 degree for 3000 hours.