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Journal Articles

Experimental study on debris bed characteristics for the sedimentation behavior of solid particles used as simulant debris

Shamsuzzaman, M.*; Horie, Tatsuro*; Fuke, Fusata*; Kamiyama, Motoki*; Morioka, Toru*; Matsumoto, Tatsuya*; Morita, Koji*; Tagami, Hirotaka; Suzuki, Toru*; Tobita, Yoshiharu

Annals of Nuclear Energy, 111, p.474 - 486, 2018/01

 Times Cited Count:9 Percentile:74.46(Nuclear Science & Technology)

Journal Articles

Preliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor

Onoda, Yuichi; Matsuba, Kenichi; Tobita, Yoshiharu; Suzuki, Toru

Mechanical Engineering Journal (Internet), 4(3), p.16-00597_1 - 16-00597_14, 2017/06

Journal Articles

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 2; Assessment of PAMR/PAHR phase in ULOF

Sogabe, Joji; Suzuki, Toru; Wada, Yusaku; Tobita, Yoshiharu

Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00393_1 - 16-00393_10, 2017/04

The achievement of In-Vessel Retention (IVR) of the accident consequences in an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, is effective and rational approach in enhancing the safety characteristics of sodium-cooled fast reactor. The objective of the present study is to show that the decay heat generated from the relocated fuels would be stably removed in post-accident-material-relocation/post-accident-heat-removal (PAMR/PAHR) phase, where the relocated fuels mean fuel discharged from the core into the low-pressure plenum through control-rod guide tubes, and fuel remnant in the disrupted core region (non-discharged fuel). As a result of the present assessments, it should be concluded that the stable cooling of the relocated fuels was confirmed and the prospect of IVR was obtained.

Journal Articles

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 1; Overview of IVR evaluation in Anticipated Transient without Scram (ATWS)

Suzuki, Toru; Sogabe, Joji; Tobita, Yoshiharu; Sakai, Takaaki*; Nakai, Ryodai

Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00395_1 - 16-00395_9, 2017/04

no abstracts in English

Journal Articles

Irradiation test of semiconductors components on the shelf for nuclear robots based on Fukushima Accidents

Kawatsuma, Shinji; Nakai, Koji; Suzuki, Yoshiharu; Kase, Takeshi

QST-M-2; QST Takasaki Annual Report 2015, P. 81, 2016/12

Radiation Tolerance of semiconductor components on the shelf, utilized on the robots for emergency response or decommissioning in nuclear facilities, should be estimated. Just after the Fukushima Daiichi NPPs accidents occurred, a guideline, of irradiation tolerance estimation and management method of semiconductor components on the shelf, was tried to be made based on the old database developed in the course of Bilateral Servo Manipulator under the high radiation and high contamination environments. The estimation was conservative, because the data in the database were old and mainly based on the test results of silicon semiconductors. Ga-As Semiconductors are coming major recently, and expected to be higher radiation tolerance. For those reason, present semiconductor devices have irradiated and the irradiation tolerance have estimated.

Journal Articles

In-vessel retention of unprotected accident in a fast reactor; Assessment of material-relocation and heat-removal behavior in ULOF

Sogabe, Joji; Suzuki, Toru; Wada, Yusaku; Tobita, Yoshiharu

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11

Journal Articles

Improvements to the simmer code model for steel wall failure based on EAGLE-1 test results

Toyoka, Junichi; Kamiyama, Kenji; Tobita, Yoshiharu; Suzuki, Toru

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11

Journal Articles

An Empirical correlation to predict the distance for fragmentation of simulated Molten-Core materials discharged into a sodium pool

Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 8 Pages, 2016/10

In order to evaluate the distance for fragmentation of molten core material discharged into the lower sodium plenum during core disruptive accidents in sodium-cooled fast reactors, experiments with simulated molten materials and coolants (water, sodium) was carried out, where an empirical correlation of the distance for fragmentation was developed. The empirical correlation developed by this study showed a good agreement with the measurement results obtained by the present experiments. It was found that in order to well-predict the distance for fragmentation in sodium, thermal phenomena, such as sodium boiling and resultant vapor expansion, needed to be considered.

Journal Articles

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 2; Assessment of PAMR/PAHR phase in ULOF

Sogabe, Joji; Suzuki, Toru; Wada, Yusaku; Tobita, Yoshiharu

Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 3 Pages, 2016/06

no abstracts in English

Journal Articles

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 1; Overview of IVR evaluation in Anticipated Transient without Scram (ATWS)

Suzuki, Toru; Sogabe, Joji; Tobita, Yoshiharu; Sakai, Takaaki*; Nakai, Ryodai

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

Journal Articles

Preliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor

Onoda, Yuichi; Matsuba, Kenichi; Tobita, Yoshiharu; Suzuki, Toru

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 10 Pages, 2016/06

Journal Articles

Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Kenichi; Suzuki, Toru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji*; Guo, L.*; Zhang, B.*

Journal of Nuclear Science and Technology, 53(5), p.698 - 706, 2016/05

AA2015-0794.pdf:2.46MB

 Times Cited Count:13 Percentile:80.58(Nuclear Science & Technology)

The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior.

Journal Articles

Lithium intercalation and structural changes at the LiCoO$$_{2}$$ surface under high voltage battery operation

Taminato, So*; Hirayama, Masaaki*; Suzuki, Kota*; Tamura, Kazuhisa; Minato, Taketoshi*; Arai, Hajime*; Uchimoto, Yoshiharu*; Ogumi, Zempachi*; Kanno, Ryoji*

Journal of Power Sources, 307, p.599 - 603, 2016/03

 Times Cited Count:28 Percentile:73.05(Chemistry, Physical)

An epitaxial-film model electrode of LiCoO$$_{2}$$(104) was fabricated on SrRuO$$_{3}$$(100)/Nb:SrTiO$$_{3}$$(100) using pulsed laser deposition. The 50 nm thick LiCoO$$_{2}$$(104) film exhibited lithium (de-)intercalation activity with a first discharge capacity of 119 mAh g$$^{-1}$$ between 3.0 and 4.4 V, followed by a gradual capacity fading with subsequent charge-discharge cycles. In contrast, a 3.2 nm thick Li$$_{3}$$PO$$_{4}$$-coated film exhibited a higher intercalation capacity of 148 mAh g$$^{-1}$$ with superior cycle retention than the uncoated film. In situ surface X-ray diffraction measurements revealed a small lattice change at the coated surface during the (de-)intercalation processes compared to the uncoated surface. The surface modification of LiCoO$$_{2}$$ by the Li$$_{3}$$PO$$_{4}$$ coating could lead to improvement of the structural stability at the surface region during lithium (de-)intercalation at high voltage.

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

A Numerical study on local fuel-coolant interactions in a simulated molten fuel pool using the SIMMER-III code

Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Annals of Nuclear Energy, 85, p.740 - 752, 2015/11

 Times Cited Count:22 Percentile:91.13(Nuclear Science & Technology)

Journal Articles

Evaluation of recriticality behavior in the material-relocation phase for Japan sodium-cooled fast reactor

Suzuki, Toru; Tobita, Yoshiharu; Nakai, Ryodai

Journal of Nuclear Science and Technology, 52(11), p.1448 - 1459, 2015/11

 Times Cited Count:6 Percentile:45.07(Nuclear Science & Technology)

Journal Articles

First analysis of local fuel-coolant interactions in a molten pool by SIMMER-III using reactor materials

Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05

Journal Articles

A Preliminary evaluation of unprotected loss-of-flow accident for a prototype fast-breeder reactor

Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Matsuba, Kenichi; Ito, Kei; Ohshima, Hiroyuki

Nuclear Engineering and Technology, 47(3), p.240 - 252, 2015/04

 Times Cited Count:20 Percentile:88.22(Nuclear Science & Technology)

Journal Articles

The Effect of coolant quantity on local fuel-coolant interactions in a molten pool

Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Annals of Nuclear Energy, 75, p.20 - 25, 2015/01

 Times Cited Count:6 Percentile:51.48(Nuclear Science & Technology)

Journal Articles

SIMMER-III analyses of local fuel-coolant interactions in a simulated molten fuel pool; Effect of coolant quantity

Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Science and Technology of Nuclear Installations, 2015, p.964327_1 - 964327_14, 2015/00

 Times Cited Count:2 Percentile:20.43(Nuclear Science & Technology)

105 (Records 1-20 displayed on this page)