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Journal Articles

Decomposition behavior of gaseous ruthenium tetroxide under atmospheric conditions assuming evaporation to dryness accident of high-level liquid waste

Yoshida, Naoki; Ono, Takuya; Yoshida, Ryoichiro; Amano, Yuki; Abe, Hitoshi

Journal of Nuclear Science and Technology, 57(11), p.1256 - 1264, 2020/11

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Emphasis has been placed on the behavior of ruthenium (Ru) in the evaporation to dryness accident due to the loss of cooling functions (EDLCF) of high-level liquid waste in fuel reprocessing plants. It is because Ru would form volatile compounds such as ruthenium tetroxide (RuO$$_{4}$$) and could be released into the environment with other coexisting gasses such as nitric acid (HNO$$_{3}$$), water (H$$_{2}$$O). To contribute to the safety evaluation of this accident, we experimentally evaluated the decomposition and chemical change behavior of the gaseous RuO$$_{4}$$ (RuO$$_{4}$$(g)) under the various atmospheric conditions: temperature and composition of coexisting gasses. As a result, the behavior of the RuO$$_{4}$$(g) was diverse depending on the atmospheric conditions. In the experiments with the dry air or H$$_{2}$$O vapor, decomposition of RuO$$_{4}$$(g) was observed. In the experiment with the mixed gas which containing HNO$$_{3}$$, almost no decomposition of the RuO$$_{4}$$(g) was observed, and chemical form of the RuO$$_{4}$$(g) was retained.

Journal Articles

Material balance evaluation of pyroprocessing for minor actinide transmutation nitride fuel

Tateno, Haruka; Sato, Takumi; Tsubata, Yasuhiro; Hayashi, Hirokazu

Journal of Nuclear Science and Technology, 57(3), p.224 - 235, 2020/03

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Fuel cycle technology for the transmutation of long-lived minor actinides (MAs) using an accelerator-driven system has been developed using the double-strata fuel cycle concept. A mononitride solid solution of MAs and Pu diluted with ZrN is a prime fuel candidate for the accelerator-driven transmutation of MAs. Pyro-reprocessing is suitable for recycling the residual MAs in irradiated nitride fuel with high radiation doses and decay heat. Spent nitride fuel is anodically dissolved, and the actinides are recovered simultaneously into a liquid cadmium cathode via molten salt electrorefining. The process should be designed to achieve the target recovery yield of MAs and the acceptable impurity level of rare earths in the recovered material. We evaluated the material balance during the pyro-reprocessing of spent nitride fuel to gain important insight on the design process. We examined the effects of changing processing conditions on material flow and quantity of waste.

Journal Articles

A Review of separation processes proposed for advanced fuel cycles based on technology readiness level assessments

Baron, P.*; Cornet, S. M.*; Collins, E. D.*; DeAngelis, G.*; Del Cul, G.*; Fedorov, Y.*; Glatz, J. P.*; Ignatiev, V.*; Inoue, Tadashi*; Khaperskaya, A.*; et al.

Progress in Nuclear Energy, 117, p.103091_1 - 103091_24, 2019/11

 Times Cited Count:3 Percentile:27.1(Nuclear Science & Technology)

The results of an international review of separation processes for spent nuclear fuel (SNF) recycling in future closed fuel cycles with the evaluation of Technology Readiness Level are reported. This study was made by the Expert Group on Fuel Recycling Chemistry (EGFRC) organised by the Nuclear Energy Agency (NEA) of the Organisation for Economic Co-operation and Development (OECD). A unique feature of this study was that processes were classified according to a hierarchy of separations aimed at different elements within spent fuel (uranium; uranium-plutonium co-recovery; minor actinides; high heat generating radionuclides) and also the Head-end processes, used to prepare the SNF for chemical separation, were included. Separation processes covered both wet (hydrometallurgical) and dry (pyro-chemical) processes.

Journal Articles

Mechanism of phase transfer of uranyl ions; A Vibrational sum frequency generation spectroscopy study on solvent extraction in nuclear reprocessing

Kusaka, Ryoji; Watanabe, Masayuki

Physical Chemistry Chemical Physics, 20(47), p.29588 - 29590, 2018/12

 Times Cited Count:6 Percentile:37.89(Chemistry, Physical)

Mechanistic understanding of solvent extraction of uranyl ions (UO$$_{2}$$$$^{2+}$$) by tributyl phosphate (TBP) will help improve the technology for the treatment and disposal of spent nuclear fuels. So far, it has been believed that uranyl ions in the aqueous phase are adsorbed to a TBP-enriched organic/aqueous interface, form complexes with TBP at the interface, and are extracted into the organic phase. Here we show that uranyl-TBP complex formation does not take place at the interface using vibrational sum frequency generation (VSFG) spectroscopy and propose an alternative extraction mechanism that uranyl nitrate, UO$$_{2}$$(NO$$_{3}$$)$$_{2}$$, passes through the interface and forms the uranyl-TBP complex, UO$$_{2}$$(NO$$_{3}$$)$$_{2}$$(TBP)$$_{2}$$, in the organic phase.

Journal Articles

Security measures at nuclear fuel facilities, 2; Internal threat countermeasure in cyber-security

Kono, Soma; Yamada, Hiroyuki; Goto, Atsushi*; Yamazaki, Katsuyuki; Nakamura, Hironobu; Kitao, Takahiko

Nihon Kaku Busshitsu Kanri Gakkai Dai-39-Kai Nenji Taikai Rombunshu (Internet), 2 Pages, 2018/11

no abstracts in English

Journal Articles

Development of cement based encapsulation for low radioactive liquid waste in Tokai Reprocessing Plant

Matsushima, Ryotatsu; Sato, Fuminori; Saito, Yasuo; Atarashi, Daiki*

Proceedings of 3rd International Symposium on Cement-based Materials for Nuclear Wastes (NUWCEM 2018) (USB Flash Drive), 4 Pages, 2018/10

At TRP, LWTF was constructed as a facility for processing low radioactive liquid waste and solid waste generated at TRP, and a cold test is been carrying out. In this facility, initially, nitrate waste liquid after separation of nuclides generated with treatment of low radioactive liquid waste was to be solidified by using borate. However, at present, it is necessary to decompose the nitrate in the liquid waste to reduce the environmental burden. For the reason, as a plan to replace the nitrate with the carbonate and to make it as a cement based encapsulation, we are studying for the introduction of the facility. Currently, as a cement solidification technology development for this liquid waste, we are studying the application of cement material based on blast furnace slag (BFS) as a main component. In this report, we show the results of the test conducted on the actual scale (200 L drum can scale).

Journal Articles

Nuclear agreement to promote cooperation between Japan and the U.S.

Suda, Kazunori

Enerugi Rebyu, 38(10), p.38 - 41, 2018/09

no abstracts in English

Journal Articles

Physical property evaluation of valve seal material at analytical radioactive liquid waste storage tanks in reprocessing facility

Goto, Yuichi; Yamamoto, Masahiko; Kuno, Takehiko; Inada, Satoshi

Nippon Hozen Gakkai Dai-15-Kai Gakujutsu Koenkai Yoshishu, p.489 - 492, 2018/07

Radioactive liquid waste from the Tokai Reprocessing Facility Analytical Laboratory is temporarily stored in intermediate waste storage tank by using receiving valves. Then, the liquid waste is transferred to liquid treatment facility by using liquid feed valves. The deterioration of the gasket part of these valves (leakage of waste liquid) was confirmed in 2004. Since then, the material of gaskets was changed from polyethylene to Teflon. In 2016, the gaskets were replaced by periodical update. Therefore, physical properties of used gaskets were investigated, and the relevance between radioactive level and degradation degree was evaluated.

Journal Articles

Optimization of disposal method and scenario to reduce high level waste volume and repository footprint for HTGR

Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi; Nishihara, Tetsuo; Tsubata, Yasuhiro; Matsumura, Tatsuro

Annals of Nuclear Energy, 116, p.224 - 234, 2018/06

 Times Cited Count:1 Percentile:74.25(Nuclear Science & Technology)

Optimization of disposal method and scenario to reduce volume of High Level Waste (HLW) and the footprint in a geological repository for High Temperature Gas-cooled Reactor (HTGR) has been performed. It was found that HTGR has great advantages to reducing HLW volume and its footprint, which are high burn-up, high thermal efficiency and pin-in-block type fuel, compared with those of LWR and has potential to reduce those more in the previous study. In this study, the scenario is optimized, and the geological repository layout is designed with the horizontal emplacement based on the KBS-3H concept instead of the vertical emplacement based on KBS-3V concept employed in the previous study. As a result, for direct disposal, the repository footprint can be reduced by 20 % by employing the horizontal without change of the scenario. By extending 40 years for cooling time before disposal, the footprint can be reduced by 50 %. For disposal with reprocessing, the number of canister generation can be reduced by 20 % by extending cooling time of 1.5 years between the discharge and reprocessing. The footprint per electricity generation can be reduced by 80 % by extending 40 years before disposal. Moreover, by employing four-group partitioning technology without transmutation, the footprint can be reduced by 90 % with cooling time of 150 years.

Journal Articles

Report on activities related to fostering nuclear security culture in Tokai Reprocessing Facility

Kitao, Takahiko; Takeuchi, Yoshikatsu; Kimura, Takashi; Kojima, Junji; Shioya, Satoshi; Tasaki, Takashi; Nakamura, Hironobu

Nihon Kaku Busshitsu Kanri Gakkai Dai-38-Kai Nenji Taikai Rombunshu (Internet), 9 Pages, 2018/04

In order to ensure and strengthen nuclear security measures, an active cultivation of nuclear security culture implemented by both entire organization and individual persons is vitally essential. Tokai reprocessing facility has conducted various activities such as case study education and training, yearly posters and patrolling the site by upper-level management that all employees hold a deep rooted belief that there is a credible insider and outsider threat, and that nuclear security is important. These activities are conducted in order to establish the foundation of beliefs and attitudes of effective nuclear security culture based on the IAEA guideline. This report introduces our activities fostering nuclear security culture in Tokai reprocessing facility that the evaluation and the continuous improvement of bidirectional activities by both top-down from multiple management levels and bottom-up from individual employee, and our challenges need to be worked on for the future.

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

Study on hydrogen generation from cement solidified products loading low-radioactive liquid wastes at Tokai Reprocessing Plant

Ito, Yoshiyuki; Matsushima, Ryotatsu; Sato, Fuminori

QST-M-8; QST Takasaki Annual Report 2016, P. 69, 2018/03

no abstracts in English

Journal Articles

Effect of Ru chemical form on corrosion behavior of structural materials in spent fuel reprocessing process

Sano, Yuichi; Ambai, Hiromu; Takahatake, Yoko

Aichi Shinkurotoron Hikari Senta 2017-Nendo Kokyoto Riyo Seika Hokokusho (Internet), 1 Pages, 2018/00

In order to elucidate the mechanism of corrosion in the reprocessing process and propose a method for suppressing corrosion, the effect of coexisting substances on the chemical form of Ru in nitric acid solution containing seawater components was evaluated. The result of XAFS measurement for Ru showed the structural change around a Ru atom due to the interaction with chloride ion, which will suppress the corrosion promoting action of Ru in nitric acid solution.

Journal Articles

Current status of research for the accident of evaporation to dryness caused by boiling of reprocessed high level radioactive liquid waste

Tamaki, Hitoshi; Yoshida, Kazuo; Abe, Hitoshi; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 9 Pages, 2017/11

An accident of evaporation to dryness caused by boiling of high level radioactive liquid waste (HLLW) is postulated as one of severe accidents caused by the loss of cooling function at the fuel reprocessing plant. This accident can be divided into early boiling stage, late boiling stage and dry-out stage by characteristics of accident evolution. It is important to estimate the amount of fission product (FP) transport between the liquid and gas phases, and the amount of FP deposition on the walls in each stage in order to estimate the release amount of FP to the environment. Various research activities have been carried out for this issue. This paper reviews these activities and presents the recent activities at JAEA for development of simulation code for this type of accident.

Journal Articles

Replacement of the glove port equipped with glove box in Nuclear Fuel Reprocessing Facility

Horigome, Kazushi; Taguchi, Shigeo; Nishida, Naoki; Goto, Yuichi; Inada, Satoshi; Kuno, Takehiko

Nippon Hozen Gakkai Dai-14-Kai Gakujutsu Koenkai Yoshishu, p.381 - 384, 2017/08

no abstracts in English

Journal Articles

Solvent extraction of uranium with ${it N}$,${it N}$-di(2-ethylhexyl)octanamide from nitric acid medium

Tsutsui, Nao; Ban, Yasutoshi; Sagawa, Hiroshi; Ishii, Sho; Matsumura, Tatsuro

Solvent Extraction and Ion Exchange, 35(6), p.439 - 449, 2017/08

 Times Cited Count:3 Percentile:76.63(Chemistry, Multidisciplinary)

Solvent extraction of uranium from a nitric acid medium was performed with ${it N}$,${it N}$-di(2-ethylhexyl)octanamide (DEHOA) by a single-stage batch method, and the distribution ratio equation of U(VI) was derived as $${it D}_{rm U}$$ = 1.1$$[rm NO^{-}_{3}]^{1.6}_{rm aq}[{rm DEHOA}]^{2}_{rm org}$$. Furthermore, the nitric acid distribution was also evaluated, and the distribution ratio equation $${it D}_{rm H}$$ = 0.12$$[rm H^{+}]^{0.76}_{rm aq}[{rm DEHOA_{rm Free}}]_{rm H}$$ was obtained. Batch experiments to evaluate the time dependence of U(VI) extraction and the U(VI) loading capacity of DEHOA were also performed. It was revealed that U(VI) extraction by DEHOA reached an equilibrium state within a few minutes, and the loading capacity was 0.71 mol/dm$$^{3}$$ (M) when the concentrations of DEHOA and nitric acid were 1.5 and 3.0 M, respectively.

Journal Articles

Effect of seawater on corrosion of SUS316L in HAW under $$gamma$$-ray irradiation

Ambai, Hiromu; Nishizuka, Yusuke*; Sano, Yuichi; Uchida, Naoki; Iijima, Shizuka

QST-M-2; QST Takasaki Annual Report 2015, P. 90, 2017/03

The spent fuel stored in the storage pools at the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company Holdings, Inc. is exposed with the environment containing seawater components, owing to the injection of seawater into the storage pools. Therefore, during reprocessing, it is expected that the spent fuel will be contaminated with seawater components, and the influence of seawater on reprocessing needs to be investigated. We conducted the corrosion tests of the HAW storage tanks under $$gamma$$-ray irradiation, and revealed that no significant effect of seawater components was emerged.

Journal Articles

Estimation of corrosion mechanisms from the data obtained by the reproduced experiments considering the actual environments; Maritime structures and nuclear facilities

Yamamoto, Masahiro

Zairyo To Kankyo, 66(1), p.3 - 12, 2017/01

The laboratory simulation tests which could be reproduced the corrosion reactions propagating in the actual environments were utilized to analyze the mechanism of corrosion phenomena. In this report, some results are introduced in the cases of maritime structures and nuclear facilities. Experimental apparatus was originally designed to obtain the data in high radioactive condition simulating actual plants. One is a result showing the effect of Np ion to the corrosion of stainless steel in nuclear fuel reprocessing plant. Corrosion mechanism was revealed that Np$$^{6+}$$ ion is reduced to Np$$^{5+}$$ ion by a corrosion reaction of stainless steel and then re-oxidized to Np$$^{6+}$$ ion in the bulk solution. And repetition of this cycle accelerated corrosion of stainless steel by a little amounts of Np addition in nitric acid solution. Another result is introduced that an effect of H$$_{2}$$O$$_{2}$$ created by radiolysis of cooling water at high radioactive environment in light water reactor.

Journal Articles

Simulation study of sludge precipitation in spent fuel reprocessing

Takeuchi, Masayuki; Aihara, Haruka; Nakahara, Masaumi; Tanaka, Kotaro*

Procedia Chemistry, 21, p.182 - 189, 2016/12

 Times Cited Count:1 Percentile:25.7

A simulation technology with electrolyte thermodynamic model has been developed to evaluate the precipitation behavior in reprocessing solution based on nitric acid solution. The simulation results were compared with the experiment data from non-radioactive simulated HLLW with ten elements and Pu-Zr-Mo solution, and the reliability of the thermodynamic model was verified. Most of the precipitation species was zirconium molybdate hydrate from the both data. It is demonstrated that the chemical species and amount of the precipitation calculated by thermodynamic model reflected well that of experiments. This study has shown the thermodynamic simulation model is one of the useful tools to estimate the behavior of precipitation from the reprocessing solution.

Journal Articles

Reduction on high level radioactive waste volume and geological repository footprint with high burn-up and high thermal efficiency of HTGR

Fukaya, Yuji; Nishihara, Tetsuo

Nuclear Engineering and Design, 307, p.188 - 196, 2016/10

AA2015-0894.pdf:0.58MB

 Times Cited Count:2 Percentile:69.03(Nuclear Science & Technology)

Reduction of High Level Waste (HLW) and footprint in a geological repository due to high burn-up and high thermal efficiency of High Temperature Gas-cooled Reactor (HTGR) has been investigated. A helium-cooled and graphite-moderated commercial HTGR was designed as a Gas Turbine High Temperature Reactor (GTHTR300), and the features are significantly high burn-up of approximately 120 GWd/t, high thermal efficiency around 50%, and pin-in-block type fuel. The pin-in-block type fuel was employed to reduce processed graphite volume in reprocessing, and effective waste loading method for direct disposal is proposed by applying the feature in this study. As a result, it is found that the number of canisters and its repository footprint per electricity generation can be reduced by 60% compared with LWR representative case for direct disposal because of the higher burn-up, higher thermal efficiency, less TRU generation, and effective waste loading proposed in this study for HTGR. For disposal with reprocessing, the number of canisters and its repository footprint per electricity generation can be reduced by 30% compared with LWR because of the 30% higher thermal efficiency of HTGR.

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