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JAEA Reports

Annual report on the effluent control of low level liquid waste in Nuclear Fuel Cycle Engineering Laboratories FY2020

Nakano, Masanao; Nakada, Akira; Kanai, Katsuta; Nagaoka, Mika; Koike, Yuko; Yamada, Ryohei; Kubota, Tomohiro; Yoshii, Hideki*; Otani, Kazunori*; Hiyama, Yoshinori*; et al.

JAEA-Review 2021-040, 118 Pages, 2021/12

JAEA-Review-2021-040.pdf:2.48MB

Based on the regulations (the safety regulation of Tokai Reprocessing Plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and by law of Ibaraki Prefecture), the effluent control of liquid waste discharged from the Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency has been performed. This report describes the effluent control results of the liquid waste in the fiscal year 2020. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other nuclear fuel material usage facilities were much lower than the limits authorized by the above regulations.

Journal Articles

Investigation of applicability of subchannel analysis code ASFRE on thermal hydraulics analysis in fuel assembly with inner duct structure in sodium cooled fast reactor

Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki

Proceedings of 28th International Conference on Nuclear Engineering; Nuclear Energy the Future Zero Carbon Power (ICONE 28) (Internet), 8 Pages, 2021/08

In the design study of an advanced sodium-cooled fast reactor (Advanced-SFR) in JAEA, the use of a specific fuel assembly (FA) with an inner duct structure called FAIDUS has been investigated to enhance safety of Advanced-SFR. Due to the asymmetric layout of fuel rods by the inner duct, it is necessary to estimate the temperature distribution to confirm feasibility of FAIDUS. For the FAIDUS, confirmation of validity of the numerical results by a subchannel analysis code named ASFRE was not enough because the reference data on the thermal hydraulics in FAIDUS have not been obtained by the mock-up experiment, yet. Therefore, the code-to-code comparisons with numerical results of ASFRE and those of a CFD code named SPIRAL was conducted. The applicability of ASFRE was indicated through the confirmation of the consistency of mechanism of the specific temperature and velocity distributions appearing around the inner duct between the numerical results by ASFRE and those by SPIRAL.

JAEA Reports

Annual report on the effluent control of low level liquid waste in Nuclear Fuel Cycle Engineering Laboratories FY2019

Nakano, Masanao; Fujii, Tomoko; Nagaoka, Mika; Koike, Yuko; Yamada, Ryohei; Kubota, Tomohiro; Yoshii, Hideki*; Otani, Kazunori*; Hiyama, Yoshinori*; Kikuchi, Masaaki*; et al.

JAEA-Review 2020-070, 120 Pages, 2021/02

JAEA-Review-2020-070.pdf:2.47MB

Based on the regulations (the safety regulation of Tokai Reprocessing Plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and by law of Ibaraki Prefecture), the effluent control of liquid waste discharged from the Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency has been performed. This report describes the effluent control results of the liquid waste in the fiscal year 2019. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other nuclear fuel material usage facilities were much lower than the limits authorized by the above regulations.

Journal Articles

Validation study of finite element thermal-hydraulics analysis code SPIRAL to a large-scale wire-wrapped fuel assembly at low flow rate condition

Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Gerschenfeld, A.*

Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.73 - 80, 2020/10

A finite element thermal-hydraulics simulation code SPIRAL has been developed in Japan Atomic Energy Agency (JAEA) to analyze the detailed thermal-hydraulics phenomena in a fuel assembly (FA) of Sodium-cooled Fast Reactors (SFRs). The numerical simulation of a large-scale sodium test for 91-pin bundle (GR91) at low flow rate condition was performed for the validation of SPIRAL with the hybrid k-e turbulence model to take into account the low Re number effect near the wall in the flow and temperature fields. Through the numerical simulation, specific velocity distribution affected by the buoyancy force was shown on the top of the heated region and the temperature distribution predicted by SPIRAL agreed with that measured in the experiment and the applicability of the SPIRAL to thermal-hydraulic evaluation of large-scale fuel assembly at low flow rate condition was indicated.

JAEA Reports

Annual report on the effluent control of low level liquid waste in Nuclear Fuel Cycle Engineering Laboratories FY2018

Nakano, Masanao; Fujii, Tomoko; Nagaoka, Mika; Inoue, Kazumi; Koike, Yuko; Yamada, Ryohei; Yoshii, Hideki*; Otani, Kazunori*; Hiyama, Yoshinori*; Kikuchi, Masaaki*; et al.

JAEA-Review 2019-045, 120 Pages, 2020/03

JAEA-Review-2019-045.pdf:2.54MB

Based on the regulations (the safety regulation of Tokai Reprocessing Plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and by law of Ibaraki Prefecture), the effluent control of liquid waste discharged from the Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency has been performed. This report describes the effluent control results of the liquid waste in the fiscal year 2018. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other nuclear fuel material usage facilities were much lower than the limits authorized by the above regulations.

Journal Articles

Subchannel analysis of thermal-hydraulics in a fuel assembly with inner duct structure of a sodium-cooled fast reactor

Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki

Journal of Nuclear Engineering and Radiation Science, 5(2), p.021001_1 - 021001_12, 2019/04

In the design study of an advanced loop-type sodium-cooled fast reactor in Japan, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been considered as one of the measures to enhance safety of the reactor during the core disruptive accident. In this study, thermal-hydraulics in FAIDUS was investigated with the in-house subchannel analysis code named ASFRE. Before the application to FAIDUS, applicability of ASFRE to FAs was confirmed through the numerical simulations for the experiments of simulated FA. Through the comparisons between the numerical results of thermal-hydraulic analyses of FAIDUS and a typical FA without the inner duct, it was indicated that significant asymmetric temperature distribution would not occur in FAIDUS at both high and low flow rate conditions.

JAEA Reports

Annual report on the effluent control of low level liquid waste in Nuclear Fuel Cycle Engineering Laboratories FY2017

Nakano, Masanao; Fujita, Hiroki; Mizutani, Tomoko; Nagaoka, Mika; Inoue, Kazumi; Koike, Yuko; Yamada, Ryohei; Yoshii, Hideki*; Hiyama, Yoshinori*; Otani, Kazunori*; et al.

JAEA-Review 2018-028, 120 Pages, 2019/02

JAEA-Review-2018-028.pdf:2.69MB

Based on the regulations (the safety regulation of Tokai Reprocessing Plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and by law of Ibaraki Prefecture), the effluent control of liquid waste discharged from the Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency has been performed. This report describes the effluent control results of the liquid waste in the fiscal year 2017. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other nuclear fuel material usage facilities were much lower than the limits authorized by the above regulations.

Journal Articles

Radiation imaging using a compact Compton camera inside the Fukushima Daiichi Nuclear Power Station building

Sato, Yuki; Tanifuji, Yuta; Terasaka, Yuta; Usami, Hiroshi; Kaburagi, Masaaki; Kawabata, Kuniaki; Utsugi, Wataru*; Kikuchi, Hiroyuki*; Takahira, Shiro*; Torii, Tatsuo

Journal of Nuclear Science and Technology, 55(9), p.965 - 970, 2018/09

 Times Cited Count:15 Percentile:94.46(Nuclear Science & Technology)

Journal Articles

Study on gas entrainment from unstable drifting vortexes on liquid surface

Hirakawa, Moe*; Kikuchi, Yuichiro*; Sakai, Takaaki*; Tanaka, Masaaki; Ohshima, Hiroyuki

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07

Gas entrainment (GE) from cover gas is one of key issue for Sodium-cooled fast reactors to prevent unexpected effects to core reactivity. By using a computational fluid dynamics (CFD) code, analyses have been conducted to estimate the drifting vortexes on water experiments which were generated as wake vortexes behind a plate obstacle in the circulating water channel. In this paper, the results of comparison between experiments and analyses were discussed and the gas core lengths from the surface vortexes were evaluated by using the evaluation tool named StreamViewer developed by Japan Atomic Energy Agency.

Journal Articles

Development of numerical analysis method for core thermal-hydraulics during natural circulation decay heat removal in SFR, 1; Validation of ASFRE code in estimation of radial heat transfer phenomena

Kikuchi, Norihiro; Doda, Norihiro; Hashimoto, Akihiko*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki

Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06

For the thermal-hydraulic design regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been developed by JAEA. ASFRE was applied to numerical simulations of several kinds of water and sodium experiments as its validation studies and it was confirmed that pressure drops and temperature distributions measured in the experiments can be well reproduced. To enhance safety of sodium-cooled fast reactor, it is required to evaluate thermal-hydraulics in a core during decay heat removal by natural circulation. It is necessary to estimate radial heat transfer phenomena between fuel assemblies. In this study, a numerical simulation of a 37-pin bundle sodium experiment with radial heat flux was carried out and it was confirmed that ASFRE can be qualitatively reproduced temperature distributions in a fuel assembly affected by radial heat transfer.

JAEA Reports

Annual report on the effluent control of low level liquid waste in Nuclear Fuel Cycle Engineering Laboratories FY2016

Nakano, Masanao; Fujita, Hiroki; Nagaoka, Mika; Inoue, Kazumi; Koike, Yuko; Yoshii, Hideki*; Hiyama, Yoshinori*; Otani, Kazunori*; Kikuchi, Masaaki*; Sakauchi, Nobuyuki*; et al.

JAEA-Review 2017-037, 119 Pages, 2018/03

JAEA-Review-2017-037.pdf:2.58MB

Based on the regulations (the safety regulation of Tokai Reprocessing Plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and bylaw of Ibaraki Prefecture), the effluent control of liquid waste discharged from the Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency has been performed. This report describes the effluent control results of the liquid waste in the fiscal year 2016. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other nuclear fuel material usage facilities were much lower than the limits authorized by the above regulations.

Journal Articles

Thermal-hydraulics analysis of fuel assembly with inner duct structure of a sodium-cooled fast reactor

Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki

Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2017 Koen Rombunshu (CD-ROM), 4 Pages, 2017/08

A specific fuel assembly named FAIDUS (Fuel Assembly with Inner Duct Structure) has been developed as one of the measures to enhance safety of the reactor in the core disruptive accident (CDA) in JAEA. Thermal-hydraulics evaluations in FAIDUS under various operation conditions including the CDA are required to confirm its design feasibility. Therefore, numerical simulations by using thermal-hydraulics analysis program named SPIRAL developed in JAEA are conducted to analyze the thermal-hydraulics in the FAIDUS. Through the numerical simulation in the FAIDUS under tentative rated operation condition of an Advanced SFR, it was indicated that the flow and temperature distribution in the FAIDUS showed the same tendency as that in ordinary FA and specific characteristics was not observed.

Journal Articles

Thermal-hydraulic analysis of fuel assembly with inner duct structure of an advanced loop-type sodium-cooled fast reactor using ASFRE code

Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 12 Pages, 2017/07

In the design study of an advanced loop-type SFR in JAEA, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been adopted as one of the measures to enhance safety of the reactor. Thermal-hydraulics evaluations of FAIDUS under various operation conditions are required to confirm its design feasibility. In this study, after the applicability of ASFRE to FAs was confirmed through the numerical analysis using simulated FA tests, thermal-hydraulic analyses of a FA without an inner duct and a FAIDUS were conducted. Through the numerical analyses, it was indicated that asymmetric temperature distribution in a FAIDUS would not be occurred and characteristics of the temperature distribution was almost the same as that in a FA without an inner duct. Under the low flow rate condition, it was expected that the local flow acceleration caused by the buoyancy force in a FAIDUS could bring the flow redistribution and make the temperature distribution flat.

JAEA Reports

Annual report on the effluent control of low level liquid waste in Nuclear Fuel Cycle Engineering Laboratories FY2015

Nakano, Masanao; Fujita, Hiroki; Kono, Takahiko; Nagaoka, Mika; Inoue, Kazumi; Yoshii, Hideki*; Otani, Kazunori*; Hiyama, Yoshinori*; Kikuchi, Masaaki*; Sakauchi, Nobuyuki*; et al.

JAEA-Review 2017-001, 115 Pages, 2017/03

JAEA-Review-2017-001.pdf:3.57MB

Based on the regulations (the safety regulation of Tokai reprocessing plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and bylaw of Ibaraki prefecture), the effluent control of liquid waste discharged from the Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency has been performed. This report describes the effluent control results of the liquid waste in the fiscal year 2015. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other nuclear fuel material usage facilities were much lower than the limits authorized by the above regulations.

Journal Articles

Development of thermal hydraulics analysis code ASFRE for fuel assembly of sodium-cooled fast reactor; Modification of distributed resistance model and validation analysis

Kikuchi, Norihiro; Ohshima, Hiroyuki; Tanaka, Masaaki; Hashimoto, Akihiko*

Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06

For the thermal-hydraulic design and safety assessment regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been and is continuously developed in JAEA. In the numerical simulation of ASFRE confirmed that the tendency to overestimate the maximum coolant temperature in a fuel assembly still remains. In this study, Distributed Resistance Model (DRM), which deals with wire-spacer wrap volumetric effect in subchannels on peripheral and axial directions, was modified and its calibration factor was optimized in order to improve the prediction accuracy of the maximum coolant temperature. A numerical simulation of a 37-pin bundle sodium experiment was also carried out and the result showed the validity of the modified DRM.

JAEA Reports

Annual report on the effluent control of low level liquid waste in Nuclear Fuel Cycle Engineering Laboratories FY2014

Watanabe, Hitoshi; Nakano, Masanao; Fujita, Hiroki; Kono, Takahiko; Inoue, Kazumi; Yoshii, Hideki*; Otani, Kazunori*; Hiyama, Yoshinori*; Kikuchi, Masaaki*; Sakauchi, Nobuyuki*; et al.

JAEA-Review 2015-030, 115 Pages, 2015/12

JAEA-Review-2015-030.pdf:25.28MB

Based on the regulations (the safety regulation of Tokai reprocessing plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and bylaw of Ibaraki prefecture), the effluent control of liquid waste discharged from the Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency has been performed. This report describes the effluent control results of the liquid waste in the fiscal year 2014. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other nuclear fuel material usage facilities were much lower than the limits authorized by the above regulations.

Journal Articles

Numerical analysis of flow field around simulated wire-wrapped fuel pins of fast reactor

Kikuchi, Norihiro; Ohshima, Hiroyuki; Imai, Yasutomo*; Hiyama, Tomoyuki; Nishimura, Masahiro; Tanaka, Masaaki

Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2015 Koen Rombunshu, p.179 - 180, 2015/08

In an economically improved sodium-cooled fast reactor, a narrower gap is considered among the fuel pins so as to achieve a high burn-up. Therefore, it is needed to evaluate thermal-hydraulic characteristics in case of a change of the gap geometry due to deformation of fuel pin caused by such as a swelling and a thermal bowing. For this purpose, a FEM analysis code, SPIRAL has been being developed in JAEA and the code validations using water or sodium experimental results have also being performed. In this study, a numerical analysis of a flow field around wire-wrapped fuel pins based on a 3-pin bundle water experiment was carried out as a validation study of SPIRAL. As a result, it was demonstrated that the hybrid-type turbulent model incorporated in SPIRAL has a high applicability to investigate the flow structure of the narrow gap in the fuel assembly.

JAEA Reports

Annual report on the effluent control of low level liquid waste in Nuclear Fuel Cycle Engineering Laboratories FY2010

Sumiya, Shuichi; Watanabe, Hitoshi; Nakano, Masanao; Fujita, Hiroki; Kono, Takahiko; Hiyama, Yoshinori; Yoshii, Hideki*; Kikuchi, Masaaki*; Otani, Kazunori*; Goto, Ichiro*

JAEA-Review 2012-006, 114 Pages, 2012/03

JAEA-Review-2012-006.pdf:3.36MB

Based on the regulations (the safety regulation of Tokai reprocessing plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, and the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and byelaw of Ibaraki prefecture), this report describes the effluent control results of liquid waste discharged from the JAEA's Nuclear Fuel Cycle Engineering Laboratories in the fiscal year 2010, from 1st April 2010 to 31st March 2011. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other facilities were much lower than the authorized limits of the above regulations.

JAEA Reports

Annual report on the effluent control of low level liquid waste in Nuclear Fuel Cycle Engineering Laboratories FY2009

Sumiya, Shuichi; Matsuura, Kenichi; Watanabe, Hitoshi; Nakano, Masanao; Fujita, Hiroki; Kono, Takahiko; Hiyama, Yoshinori; Yoshii, Hideki*; Fujii, Jun*; Kikuchi, Masaaki*; et al.

JAEA-Review 2010-072, 115 Pages, 2011/02

JAEA-Review-2010-072.pdf:2.1MB

Based on the regulations (the safety regulation of Tokai reprocessing plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, and the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and byelaw of Ibaraki prefecture), this report describes the effluent control results of liquid waste discharged from the JAEA's Nuclear Fuel Cycle Engineering Laboratories in the fiscal year 2009, from 1st April 2009 to 31st March 2010. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other facilities were much lower than the authorized limits of the above regulations.

JAEA Reports

Annual report of Nuclear Emergency Assistance and Training Center (April 1, 2009 - March 31, 2010)

Kanamori, Masashi; Shirakawa, Yusuke; Yamashita, Toshiyuki; Okuno, Hiroshi; Terunuma, Hiroshi; Ikeda, Takeshi; Sato, Sohei; Terakado, Naoya; Nagakura, Tomohiro; Fukumoto, Masahiro; et al.

JAEA-Review 2010-037, 60 Pages, 2010/09

JAEA-Review-2010-037.pdf:3.11MB

When a nuclear emergency occurs in Japan, the Japan Atomic Energy Agency (JAEA) provides technical support to the National government, local governments, police, fire station and license holder etc. They are designated public organizations conforming to the basic law on emergency preparedness and the basic plan for disaster countermeasures. The Nuclear Emergency Assistance & Training Center (NEAT) of JAEA provides a comprehensive range of technical support activities to an off-site center in case of a nuclear emergency. Specifically, NEAT gives technical advice and information, provides for the dispatch of specialist as required, supplies emergency equipments and materials to the national government and municipal office. NEAT provide various lectures and training course concerning nuclear disaster prevention for those personnel taking an active part in emergency response organizations at normal time. And NEAT researches on nuclear disaster prevention and also cooperate with international organizations. This annual report summarized the activities of JAEA/NEAT in the fiscal year 2009.

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