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Journal Articles

Outline of decommissioning plan of Tokai Reprocessing Plant

Okano, Masanori; Akiyama, Kazuki; Taguchi, Katsuya; Nagasato, Yoshihiko; Omori, Eiichi

Dekomisshoningu Giho, (57), p.53 - 64, 2018/03

The construction of Tokai Reprocessing Plant (TRP) was initiated in June 1971, and its hot test using spent fuel started in September 1977. Thereafter TRP had been operated to reprocess 1,140 tons of spent fuel for approximately 30 years until May 2007, according to the reprocessing contract with domestic electric power companies. JAEA announced a policy of TRP in report of JAEA reform plan published in September 2014. The policy shows that TRP will shift to a decommissioning stage by economic reasons. Based on the policy, application of approval for TRP decommissioning plan was submitted to Nuclear Regulation Authority (NRA) in June 2017. This plan provides basic guidelines such as procedures for decommissioning and specific activities for risk reduction, and implementation divisions of decommissioning, management of spent fuels and radioactive wastes, decommissioning budget, and decommissioning schedule. The process of TRP decommissioning is planned to continue for approximately 70 years until the release of controlled areas of approximately 30 facilities.

Journal Articles

Nitric acid concentration dependence of dicesium plutonium(IV) nitrate formation during solution growth of uranyl nitrate hexahydrate

Nakahara, Masaumi; Kaji, Naoya; Yano, Kimihiko; Shibata, Atsuhiro; Takeuchi, Masayuki; Okano, Masanori; Kuno, Takehiko

Journal of Chemical Engineering of Japan, 46(1), p.56 - 62, 2013/01

 Times Cited Count:1 Percentile:90.56(Engineering, Chemical)

The influence of HNO$$_{3}$$ concentration in the solution on the formation of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ was evaluated in the U crystallization process. The solubility of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ in a uranyl nitrate solution was found to decrease with increasing HNO$$_{3}$$ concentration in the solution. In the U crystallization experiments with the dissolver solution of irradiated fast reactor fuel, Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ formed with 6.5 mol/dm$$^{3}$$ HNO$$_{3}$$ concentration in the mother liquor, and the decontamination factor of Cs for the uranyl nitrate hexahydrate crystals was low. Meanwhile, Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ did not precipitate with uranyl nitrate hexahydrate crystals under the condition of 4.0 mol/dm$$^{3}$$ HNO$$_{3}$$ concentration in the mother liquor, and Cs could be separated from the uranyl nitrate hexahydrate crystals.

Journal Articles

Measurement of $$^{89}$$Sr and $$^{90}$$Sr in water accumulated in turbine building at Fukushima Daiichi

Asai, Shiho; Okano, Masanori; Kameo, Yutaka

Hosha Kagaku Nyusu, (25), p.25 - 28, 2012/03

no abstracts in English

Journal Articles

Characteristics of dicesium plutonium(IV) nitrate formation in separation system of uranyl nitrate hexahydrate crystal

Nakahara, Masaumi; Yano, Kimihiko; Shibata, Atsuhiro; Takeuchi, Masayuki; Okano, Masanori; Kuno, Takehiko

Procedia Chemistry, 7, p.282 - 287, 2012/00

 Times Cited Count:1 Percentile:27.97

For decontamination of Cs and Pu compound, Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$, precipitated in the U cooling crystallization method, solubility measurement of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ in a uranyl nitrate solution and a U crystallization experiments were carried out with the dissolver solution derived from irradiated fast neutron reactor core fuel. The solubility of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ in the uranyl nitrate solution decreased with decreasing temperature. In the crystallization experiments, the decontamination factors of Cs and Pu for uranyl nitrate hexahydrate crystal decrease with increasing the Cs concentration in the feed solution because Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ formed in the course of U crystallization. Basic data were obtained for the formation behavior of Cs$$_{2}$$Pu(NO$$_{3}$$)$$_{6}$$ in the U crystallization process.

Journal Articles

Measurement of isotopic composition of lanthanides in reprocessing process solutions by high-performance liquid chromatography with inductively coupled plasma mass spectrometry (HPLC/ICP-MS)

Okano, Masanori; Jitsukata, Shu*; Kuno, Takehiko; Yamada, Keiji

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12

Journal Articles

Determinations of plutonium and curium in the insoluble materials of spent fuel dissolver solutions at the Tokai Reprocessing Plant

Okano, Masanori; Kuno, Takehiko; Nemoto, Hirokazu*; Yamada, Keiji; Watahiki, Masaru; Hiyama, Toshiaki

Proceedings of INMM 50th Annual Meeting (CD-ROM), 9 Pages, 2009/07

no abstracts in English

JAEA Reports

Evaluation technology for burnup and generated amount of plutonium by measurement of xenon isotopic ratio in dissolver off-gas at reprocessing facility (Joint research)

Okano, Masanori; Kuno, Takehiko; Takahashi, Ichiro*; Shirozu, Hidetomo; Charlton, W. S.*; Wells, C. A.*; Hemberger, P. H.*; Yamada, Keiji; Sakai, Toshio

JAEA-Technology 2006-055, 38 Pages, 2006/12

JAEA-Technology-2006-055.pdf:3.33MB

The amount of Pu in the spent fuel was evaluated from Xe isotopic ratio in off-gas in reprocessing facility, is related to burnup. Six batches of dissolver off-gas at spent fuel dissolution process were sampled from the main stack in Tokai Reprocessing Plant during BWR fuel reprocessing campaign. Xenon isotopic ratio was determined with GC/MS. Burnup and generated amount of Pu were evaluated with Noble Gas Environmental Monitoring Application code (NOVA), developed by Los Alamos National Laboratory. Inferred burnup evaluated by Xe isotopic measurements and NOVA were in good agreement with those of the declared burnup in the range from -3.8% to 7.1%. Also, the inferred amount of Pu in spent fuel was in good agreed with those of the declared amount of Pu calculated by ORIGEN code in the range from -0.9% to 4.7%. The evaluation technique is applicable for both burnup credit to achieve efficient criticality safety control and a new measurement method for safeguards inspection.

Journal Articles

Composition of insoluble materials in highly active liquid waste at Tokai Reprocessing Plant

Kuno, Takehiko; Nakamura, Yoshinobu; Okano, Masanori; Sato, Soichi; Watahiki, Masaru

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 3 Pages, 2005/10

The elemental composition and amount of insoluble materials in highly active liquid waste (HALW) were measured at the Tokai Reprocessing Plant. To evaluate the composition of the insoluble material between the treatment and storage processes, samples from the evaporator and two vessels which differ in the storage period, were taken. The concentration of insoluble material evaluated by both its weight on filter paper after filtration as well as by filtration volume showed no difference for three different samples. Inductively coupled plasma atomic emission spectrometry and isotope dilution mass spectrometry were used to determine the elements in samples dissolved by sulfate fusion. Analytical data revealed that Zr and Mo were the main components of the insoluble material in three samples, but low levels of Rh, Ru and Pd were also present. These results suggest that most of insoluble material had re-precipitated from the HALW solution during the concentration process.

Journal Articles

Development of Analysis Method for Plutonium Amount and Burn up by Measurement of Xenon Isotopic Ratio in Dissolver Off-Gas at Reprocessing Facility

Okano, Masanori; Kuno, Takehiko; Charlton, W. S.*; Wells, C. A.*; Hemberger, P. H.*

46th Annual Meeting of the INMM, Final Program, Abstract 352, 0 Pages, 2005/07

According to achieve the International Atomic Energy Agency's safeguards criteria, development of simple and rapid determination method of Pu amount in the spent fuel has been required. It is known that stable xenon isotopes produced with fission reflect spent fuel characteristics (fuel type, burn up, Pu amount). Since dissolver off-gas which is released with dissolution of spent fuel in reprocessing facility, contains most of gaseous fission products, xenon isotopic ratio in dissolver off-gas sampled from the stack would be expected to provide a new measurement method for application to safeguards inspection. We have studied a new technique to evaluate Pu isotopic ratio and its amount in the fuel from Xe isotopic ratio, which is related to burn up. The technique is also applicable for evaluating burn up credit to achieve efficient critical safety control. Six batches of off gas from dissolution process, DOG: Dissolver off gas, were obtained with press pump from the stack in Tokai Reprocessing Plant at BWR fuel (approx. 30GWD/MTU) reprocessing campaign. Xe isotopic ratio was determined with Gas Chromatography- Quadrupole Mass Spectrometer, GC-QMS. Xe concentrations in the DOGs sampled at the stack were 1,000 to 10,000 ppm, which were enough concentrations for determining minor Xe isotopes. We evaluate burn up and Pu amount with NOVA, Noble Gas Environmental Monitoring Application code, developed by Los Alamos National Laboratory. Inferred burn up was realized 94% of confidences level, which is the conventional definition based on the standard deviation of the results. We also calculated Pu amount in the spent fuel by ORIGEN code based on declared burn up. Based on the two different evaluations for Pu amount, we found that there is 7% difference between NOVA and ORIGEN code. It is summarized that there is possibility of developed method being used newly burn up monitoring and safeguards technique.

Journal Articles

Development of Analysis Method for Plutonium amount and Burn up by Measurement of Xenon Isotopic Ratio in Dissolver Off-Gas at Reprocessing Facility

Okano, Masanori; Kuno, Takehiko; Takahashi, Ichiro*; Charlton, W. S.*; Wells, C. A.*; P. H., Hemberger,*

46th Annual Meeting of the INMM, Final Program, Abstract 352, 0 Pages, 2005/07

We have studied a new technique to evaluate Pu amount in the spent fuel from Xe isotopic ratio in off-gas in reprocessing facility, which is related to burn up. The technique is applicable for both burn up credit to achieve efficient criticality safety control and a new measurement method for safeguards inspection. Six batches of dissolver off-gas (DOG) at spent fuel dissolution process were sampled with press pump from the main stack in Tokai Reprocessing Plant (TRP) during BWR fuel (approx. 30GWD/MTU) reprocessing campaign. Xe isotopic ratio was determined with Gas Chromatography-Quadrupole Mass Spectrometer (GC-QMS). Burn up and Pu amount were evaluated with NOVA, Noble Gas Environmental Monitoring Application code, developed by Los Alamos National Laboratory. Agreement between declared burn up and inferred burn up evaluated by Xe isotopic measurements and NOVA was at the 94% confidence level. We found that Pu amount in spent fuel evaluated by developed method agreed with declared value calculated by ORIGEN code in the range of -0.9% to 4.7%.

JAEA Reports

Feasibility Study on Commercialization of Fast Breeder Reactor Cycle Systems Interim Report of Phase II; Technical Study Report for Reactor Plant Systems

Konomura, Mamoru; Ogawa, Takashi; Okano, Yasushi; Yamaguchi, Hiroyuki; Murakami, Tsutomu; Takaki, Naoyuki; Nishiguchi, Youhei; Sugino, Kazuteru; Naganuma, Masayuki; Hishida, Masahiko; et al.

JNC-TN9400 2004-035, 2071 Pages, 2004/06

JNC-TN9400-2004-035.pdf:76.42MB

The attractive concepts for Sodium-, lead-bismuth-, helium- and water-cooled FBRs have been created through using typical plant features and employing advanced technologies. Efforts on evaluating technological prospects of feasibility have been paid for these concepts. Also, it was comfirmed if these concepts satisfy design requierments of capability and performance presumed in the feasibilty study on commertialization of Fast Breeder Reactor Systems. As results, it was concluded that the selection of sodium-cooled reactor was most rational for practical use of FBR technologies in 2015.

Journal Articles

Determination of Uranium, Curium and Plutonium in the Hulls

Kuno, Takehiko; Okano, Masanori; Sato, Soichi; *; Jitsukata, Shu*

Final Program Abst. P.60, 60 Pages, 2003/00

Oral presentation

Composition of insoluble materials in highly active liquid waste at Tokai Reprocessing Plant

Okano, Masanori; Kuno, Takehiko; Watahiki, Masaru; Yamada, Keiji; Kurakata, Koichiro

no journal, , 

no abstracts in English

Oral presentation

Evaluation technology for burnup and generated amount of plutonium by measurement of xenon isotopic ratio in dissolver off-gas at reprocessing facility

Okano, Masanori; Kuno, Takehiko; Yamada, Keiji; Sakai, Toshio

no journal, , 

no abstracts in English

Oral presentation

Determination of Pu/Cm ratio in sludge at Tokai reprocessing plant

Okano, Masanori; Nemoto, Hirokazu*; Jitsukata, Shu*; Yamada, Keiji; Sakai, Toshio

no journal, , 

no abstracts in English

Oral presentation

Determination of plutonium in insoluble residue at reprocessing facility

Igarashi, Kazuto*; Nemoto, Hirokazu*; Okano, Masanori; Yamada, Keiji; Sakai, Toshio

no journal, , 

no abstracts in English

Oral presentation

Determination of impurity in dissolved solution of spent nuclear fuel by ICP-AES

Okano, Masanori; Igarashi, Kazuto*; Nemoto, Hirokazu*; Yamada, Keiji; Sakai, Toshio

no journal, , 

no abstracts in English

Oral presentation

Present status of BA DEMO design activity

Tobita, Kenji; Nishio, Satoshi; Nishitani, Takeo; Ozeki, Takahisa; Araki, Masanori; Okano, Kunihiko*; Hiwatari, Ryoji*; Ogawa, Yuichi*

no journal, , 

During the first three years of BA DEMO design activity, exchange of opinions is carried out between Japanese and European experts in a workshop style. In the previous two workshops, the both parties were devoted to discussions on definition of DEMO, role of DEMO in their fusion development program, requirements for DEMO and issues on DEMO physics and engineering. Throughout the discussion, common design issues which are not independent DEMO design concepts, including (1) divertor, (2) maintenance, (3) superconducting magnet, (4)current drive (steady state operation), etc. Key issues regarding each subject are removal of high heat load in the divertor under severe neutron environment, maintenance scheme which could provide high availability, need for high field superconducting magnet, and current drive schemes favorable in the aspects of the current drive efficiency and the controllability of current profile.

Oral presentation

Determination of plutonium and curium in insoluble residue of spent fuel dissolver solution by alpha spectrometry

Okano, Masanori; Kuno, Takehiko; Nemoto, Hirokazu*; Igarashi, Kazuto*; Yamada, Keiji; Watahiki, Masaru

no journal, , 

no abstracts in English

Oral presentation

Determination of uranium and acidity in reprocessing process solutions by immersion spectroscopic probe

Kuno, Takehiko; Nemoto, Hirokazu*; Okano, Masanori; Igarashi, Kazuto*; Yamada, Keiji; Watahiki, Masaru

no journal, , 

no abstracts in English

30 (Records 1-20 displayed on this page)