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COda, Chie; Walker, C.; Chino, Daisuke*; Ichige, Satoru; Honda, Akira; Sato, Tsutomu*; Yoneda, Tetsuro*
Applied Clay Science, 93-94, p.62 - 71, 2014/05
Times Cited Count：7 Percentile：24.59(Chemistry, Physical)Na-montmorillonite dissolution in a 0.3M NaOH solution has been investigated at pH12 and 70C. The flow-through dissolution experiments were conducted in a dispersed system with varying concentrations of Si and Al to derive a Na-montmorillonite dissolution rate, as a non-linear function of the Gibbs free energy of reaction, dGr. This rate equation was used to simulate the batch-type Na-montmorillonite reaction experiments conducted in a coagulated system. The model simulation of the batch-type experiment adopting the empirical rate equations of Na-montmorillonite dissolution and secondary mineral analcime precipitation were able to reproduce the measured changes in the amount of dissolved Na-montmorillonite and concentrations of Si and Al in solution. The results showed that the empirical rate equation of Na-montmorillonite dissolution determined in the dispersed system was applicable to the coagulated system over a higher dGr range and that the concentrations of Si and Al in the batch experiment were controlled by the precipitation of analcime.
Kamei, Gento; Honda, Akira; Oda, Chie; Hirano, Fumio; Ichige, Satoru; Kurimoto, Yoshitaka; Hoshino, Seiichi; Akagi, Yosuke; Sato, Nobuyuki; Takahashi, Kuniaki; et al.
JAEA-Research 2012-010, 80 Pages, 2012/06
JAEA-Research-2012-010.pdf:7.45MB
Based on Japanese governmental policy and general scheme, research and development of geological disposal technology for TRU waste has been proceeding to improve reliability of the safety assessment of the co-locational disposal of TRU waste and of HLW, to expand the basement of generic safety assessment, and to develop the alternative technology to cope with the broad geologic environment of Japan. Japan Atomic Energy Agency is dealing with the assignments in the governmental generic scheme. We report here the progress of the studies at the end of H22 (2010) Japanese fiscal year and their products during the last 5 years. These include (1) evaluation of long-term mechanical stability in the near-field including development of a creep mode of rock and analyses of mechanical behavior of TRU waste repository, (2) performance assessment of the disposal system including cementitious material alteration, bentonite and hostrock alteration with alkaline solution and nitrate effect, and (3) alternative technology development including decomposition of nitrate.
Kamei, Gento; Honda, Akira; Mihara, Morihiro; Oda, Chie; Ichige, Satoru; Kurimoto, Yoshitaka; Hoshino, Seiichi; Akagi, Yosuke; Sato, Nobuyuki; Murakami, Hiroshi*; et al.
JAEA-Research 2011-002, 82 Pages, 2011/03
JAEA-Research-2011-002.pdf:5.64MB
Based on Japanese governmental policy and general scheme, research and development of geological disposal technology for TRU waste has been proceeding to improve reliability of the safety assessment of the co-locational disposal of TRU waste and of HLW, to expand the basement of generic safety assessment, and to develop the alternative technology to cope with the broad geologic environment of Japan. Japan Atomic Energy Agency is dealing with the assignments in the governmental generic scheme. We report here the progress of the studies at the end of H20 (2008) Japanese fiscal year, which are (1) evaluation of long-term mechanical stability in the near-field including development of a creep mode of rock and analyses of mechanical behavior of TRU waste repository, (2) performance assessment of the disposal system including data acquisition and preparation on radionuclides migration, cementitious material alteration, bentonite and hostrock alteration with alkaline solution and nitrate effect, and (3) alternative technology development including decomposition of nitrate.
Ichige, Satoru; Yamaguchi, Kohei; Oda, Chie
JAEA-Technology 2009-065, 120 Pages, 2010/02
JAEA-Technology-2009-065.pdf:10.65MB
The focus of the present study was to examine the alteration of bentonite in high pH saline groundwaters. Two solutions were used in batch immersion experiments of bentonite. The first solution was prepared using a mixture of NaOH and NaCl (NN), and the second solution was prepared using synthetic Region 1 water (high K and Na content) and synthetic seawater (SR). Analysis showed that bentonite altered to analcime in the NN solution and to analcime and phillipsite-K in the SR solution. Moreover, the generation of calcium silicate hydrate and calcium aluminosilicate hydrate were extrapolated in the SR solution based on the concentrations of dissolved species. These alteration products were in accord with Oda et al. (2005), who summarized the possible relationships between the secondary mineral assemblage of bentonite under high pH conditions and the influence of solution composition.
Kamei, Gento; Honda, Akira; Mihara, Morihiro; Oda, Chie; Murakami, Hiroshi; Masuda, Kenta; Yamaguchi, Kohei; Matsuda, Setsuro; Ichige, Satoru; Takahashi, Kuniaki; et al.
JAEA-Research 2009-046, 80 Pages, 2010/01
JAEA-Research-2009-046.pdf:9.1MB
Based on Japanese governmental policy and general scheme, research and development of geological disposal technology for TRU waste has been proceeding to improve reliability of the safety assessment of the co-locational disposal of TRU waste and of HLW, to expand the basement of generic safety assessment, and to develop the alternative technology to cope with the broad geologic environment of Japan. Japan Atomic Energy Agency is dealing with the assignments in the governmental generic scheme. We report here the progress of the studies at the end of H20 (2008) Japanese fiscal year.
Kamei, Gento; Honda, Akira; Mihara, Morihiro; Oda, Chie; Murakami, Hiroshi; Masuda, Kenta; Yamaguchi, Kohei; Matsuda, Setsuro; Ichige, Satoru; Takahashi, Kuniaki; et al.
JAEA-Research 2008-082, 84 Pages, 2008/11
JAEA-Research-2008-082.pdf:2.52MBJAEA-Research-2008-082(errata).pdf:0.15MB
Based on Japanese governmental policy and general scheme, research and development of geological disposal technology for TRU waste has been proceeding to improve reliability of the safety assessment of the co-locational disposal of TRU waste and of HLW, to expand the basement of generic safety assessment, and to develop the alternative technology to cope with the broad geologic environment of Japan. Japan Atomic Energy Agency is dealing with the assignments in the governmental generic scheme. We report here the progress of the studies at the end of H19 (2007) fiscal year, which are (1) treatment and packaging of TRU waste including applicability of calcination for unpacking and sorting of wastes, characterization and inspection methodology of TRU waste, (2) mechanical assessment for the near-field structure including model development and preparation, introduction of hostrock creep model and coupling analysis of deformation of hostrock and engineered barrier. (3) performance assessment of the disposal system including data acquisition and preparation on radionuclides migration, cementitious material alteration, bentonite and hostrock alteration with alkaline solution and nitrate effect, and (4) alternative technology development including decomposition of nitrate.
Kamei, Gento; Honda, Akira; Mihara, Morihiro; Oda, Chie; Murakami, Hiroshi; Masuda, Kenta; Yamaguchi, Kohei; Nakanishi, Hiroshi*; Sasaki, Ryoichi*; Ichige, Satoru*; et al.
JAEA-Research 2007-067, 130 Pages, 2007/09
JAEA-Research-2007-067.pdf:28.78MB
After the publication of the 2nd progress report of geological disposal of TRU waste in Japan, policy and general scheme of future study for the waste disposal in Japan was published by ANRE and JAEA. This annual report summarized aim and progress of individual problem, which was assigned into JAEA in the published policy and general scheme. The problems are as follows; characteristics of TRU waste and its geologic disposal, treatment and waste production, quality control and inspection methodology for waste, mechanical analysis of near-field, data acquisition and preparation on radionuclides migration, cementitious material transition, bentonite and rock alteration in alkaline solution, nitrate effect, performance assessment of the disposal system and decomposition of nitrate as an alternative technology.
Ichige, Satoru*; Honda, Akira
JNC TN8430 2005-003, 54 Pages, 2005/09
JNC-TN8430-2005-003.pdf:2.54MB
Bentonite clay may be used as a barrier to retard the migration of radioactive species by sorption ability and low permeability. Also, the cemetitious material is envisaged as the structural and backfilling material in the vaults. The groundwater in contact with the cementitious material will be alkaline condition in the repository and this alkaline condition will affect the performance of the bentonite. Therefore, it is necessary to investigate the altered bentonite for estimating performance of the in disposal system. In this study, to investigate interaction of amount of montmorillonite and swelling power of altered bentonite, the swelling experiment of altered bentonite were carried out. The altered bentonite were made using synthetic cement leachates (pH=14) at high temperatures (160 deg). The block sample heated for 10 days and 17 days, and the powder sample heated for 7 days. Three samples were carried out swelling experiment and the amounts of montmorillonite were determined by Methylene Blue adsorption method. It was shown that swelling power and quantity of montmorillonit have dependence. The volumes of the block sample which heated for ten days increased in a swelling experiment slowly. The cementation that occurred by the process that making altered bentonite came loose slowly, and it seems that a sample swelled. It is for future work to investigate relation with swelling power and Amount of montmorillonite. Furthermore, it is to investigate quantity of montmorillonite which there is by cementation loosely.
Sumino, Kozo; Ashida, Takashi; Kawahara, Hirotaka; Ichige, Satoshi; Isozaki, Kazunori; Nakai, Satoru
Proceedings of Operating Nuclear Facility Safety(2004ONFS),p204-216, p.204 - 216, 2004/11
None
Kawahara, Hirotaka; Isozaki, Kazunori; Ishii, Takayuki; Ichige, Satoshi; Nose, Shoichi; Sakaba, Hideo; Nakai, Satoru
JNC TN9410 2004-016, 106 Pages, 2004/06
JNC-TN9410-2004-016.pdf:8.47MB
A key part of the upgrade of the experimental fast reactor JOYO to the MK-III design was the replacement of the dump heat exchangers. MK-III function tests (SKS-1) of the new dump heat exchangers were carried out from August 27,2001 through September 13,2001. The major results of the function tests of the dump heat exchangers were as follows: (1) Air flow of the main blower with an inlet vane opening of 50% was confirmed to exceed the design rated flow of 7,700m3/min. It was also demonstrated that an inlet vane opening of 100% provides about 130% of the design rated flow. This is because the new DHX flow route has more low pressure loss than the design value. (2) Tests of the air flow of the main blower demonstrated that with a fully opened inlet damper a full opened outlet damper and an inlet vane opening of O% provides about 5% of the design rated flow. (3) Free flow coast down characteristics of the main blower achieved an inlet vane O% opening in an average of 7.9 seconds. Revolutions per minute of the main blower reached zero in an average of 8.7 seconds. The delay time from the opening of the vacuum contact breaker to the air flow decrease was approximately 1 second. This was a more conservative value than the 5 seconds assumed in design thermal transient analyses. (4) The loudest noise occurred with the main blower operating with a 25% inlet vane opening. At that time, the noise around the main blower was approximately 100dB, and in the surrounding monitoring area boundary, the noise was 50dB. This was confirmed to be within the standard of the Ibaraki prefectural ordinance. (5) Although the MK-III inlet vane and inlet damper drive unit was bigger than the MK-II unit, the accumulator tank was confirmed to provide sufficient volume during a compression air loss event.
Sakasai, Akira; Ishida, Shinichi; Matsukawa, Makoto; Akino, Noboru; Ando, Toshinari*; Arai, Takashi; Ezato, Koichiro; Hamada, Kazuya; Ichige, Hisashi; Isono, Takaaki; et al.
Nuclear Fusion, 44(2), p.329 - 334, 2004/02
no abstracts in English
Sakasai, Akira; Ishida, Shinichi; Matsukawa, Makoto; Akino, Noboru; Ando, Toshinari*; Arai, Takashi; Ezato, Koichiro; Hamada, Kazuya; Ichige, Hisashi; Isono, Takaaki; et al.
Nuclear Fusion, 44(2), p.329 - 334, 2004/02
Times Cited Count：7 Percentile：22.95(Physics, Fluids & Plasmas)no abstracts in English
Kawahara, Hirotaka; Kawahara, Hirotaka; Ichige, Satoshi; Isozaki, Kazunori; Nakai, Satoru
Proceedings of 12th International Conference on Nuclear Engineering (ICONE-12) (CD-ROM), 0 Pages, 2004/00
A recently completed major upgrade of the JOYO experimental sodium-cooled fast reactor, to the MK-III design, increased its irradiation capability approximately four times. 0ne major change was a 40% increase in thermal power to 140 MWt, which necessitated the replacement of the heat exchangers. Each of the two coolant loops includes an intermediate heat exchanger (IHX) and sodium pump in the primary system, and two dump heat exchangers (DHXs) and a pump in the secondary system. The heat transfer area of the finned tubes in each (air-cooled) DHX was doubled, compared to the old design, to achieve a 35 MWt rating, Major challenges in the replacement of secondary components, such as piping and DHX, were control of impurity ingress into the sodium system, and integrity assurance of the welding. Damage to existing components and systems was avoided during cutting and welding operations by taking measures to Prevent ingress of air into the sodium systems. The measures included use of seal b
Ishii, Takayuki; Isozaki, Kazunori; Ashida, Takashi; Minakawa, Satoru; Terakado, Tsuguo; Nogami, Hiroshi*; Kakurai, Katsuhiko*; Ueda, Soji*; Kawahara, Hirotaka; Ichige, Satoshi; et al.
JNC TN9410 2002-013, 86 Pages, 2002/11
JNC-TN9410-2002-013.pdf:68.29MB
The MK-III project has been proceeding to improve the irradiation capability of the experimental fast reactor JOY0. The MK-III project has three major purposes such as increase high neutron flux, improvement of plant availability factor and upgrading in irradiation techniques. Thermal output by core transformation for high neutron flux was increased from 100MWt to 140MWt. The main components in the cooling system such as IHX(Intermediate Heat Exchanger) and DHX(Dump Heat Exchanger) were replaced in MK-III modification in order to increase heat removal capability. Heat transfer capacity of IHX was increased from 50MWt/Unit to 70MWt/Unit and that of DHX was increased from 25MWt/Unit to 35MWt/Unit. These components replacement has been safety completed from october 30, 2000 through September 21, 2001. This report summarizes the way, results and safety measures about dismantling of no sodium adhered components and such as DHX blower and sodium cleaning of sodium components such as the DHX, the pipes connected with DHX and secondary side of IHX. Dismantling and sodium cleaning of secondary cooling system components were performed safely and efficiently as almost planned. The total amount of removed sodium was about 13.5kg.
Kurosawa, Susumi; Shibata, Masahiro; Ueta, Shinzo*; Ichige, Satoru*; Hayashi, Kenichi*; Yui, Mikazu
Nihon Genshiryoku Gakkai Wabun Rombunshi, 1(2), p.244 - 248, 2002/00
None
Sakasai, Akira; Ishida, Shinichi; Matsukawa, Makoto; Kurita, Genichi; Akino, Noboru; Ando, Toshinari*; Arai, Takashi; Ichige, Hisashi; Kaminaga, Atsushi; Kato, Takashi; et al.
Proceedings of 19th IEEE/NPSS Symposium on Fusion Engineering (SOFE), p.221 - 225, 2002/00
no abstracts in English
Ichikawa, Nozomi*; Hamamoto, Takafumi*; Sasamoto, Hiroshi; Ichige, Satoru*; Kawakita, Ryohei; Fujisaki, Kiyoshi*
no journal, ,
For model validation of bentonite behavior under high alkaline condition, a batch type reaction experiment and modeling were performed. In the batch type reaction experiment, simulated solutions for the leachate from cement material (Region I solution of pH13 simulated by mixture of both 0.2M NaOH and KOH solutions, Region II solution of pH12.5 simulated by 0.016M CaOH
solution) and bentonite (Kunigel V1 and Kunipia F) were reacted under L/S ratio of 50 mL/g and temperature for either 25 or 50 degrees, and for 2 years as the maximum duration. As the results, for example of the experiment for reactions between Region I solution and Kunipia F at 50 degrees, the peak intensity for montmorillonite slightly decreased with time and the new formed secondary mineral suggested as phillipsite occur. Geochemical modeling for this experimental case was performed. Regarding the amount of montmorillonite dissolution, the modeling considering the reactions of both montmorillonite dissolution and ion exchange approximated the experimental result well, however, the addition of the secondary mineral precipitation reaction was less effective. Sensitivity analyses considering the parameter uncertainties of secondary mineral precipitation reaction stimulated that the inconsistency was not dependent on the parameter uncertainties.
