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Kubo, Shigenobu; Chikazawa, Yoshitaka; Ohshima, Hiroyuki; Kamide, Hideki
Handbook of Generation IV Nuclear Reactors, Second Edition, p.173 - 194, 2023/03
Handbook of Generation IV Nuclear Reactors, Second Edition is a fully revised and updated comprehensive resource on the latest research and advances in generation IV nuclear reactor concepts. Editor Igor Pioro and his team of expert contributors have updated every chapter to reflect advances in the field since the first edition published in 2016. JAEA contributes to Chapter 5; Sodium-cooled Fast Reactors (SFRs) and Chapter 12; Generation-IV Sodium-cooled Fast Reactor (SFR) concepts in Japan. Major characteristics and current technology developments including safety enhancement were described in Chapter 5. Chapter 12 shows design activities of SFR. Innovative technology developments, and update of the Japan sodium-cooled fast reactor design with lessons learned from the TEPCO Fukushima Daiichi NPP accident.
Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05
Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. In this paper, the numerical simulation results of the preliminary analysis for the sodium experiment with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX.
Kondo, Masatoshi*; Okubo, Nariaki; Irisawa, Eriko; Komatsu, Atsushi; Ishikawa, Norito; Tanaka, Teruya*
Energy Procedia, 131, p.386 - 394, 2017/12
Times Cited Count：6 Percentile：95.14(Energy & Fuels)The chemical behaviors of lead (Pb) based coolants in the air ingress accident of fast reactors were investigated by means of the thermodynamic considerations and the static oxidation experiments for Pb alloys at various chemical compositions. The results of the static oxidation tests for lead-bismuth (Pb-Bi) alloys indicated that Pb was depleted from the alloy due to the preferential formation of PbO in air at 773K. Pb-Bi oxide and Bi
O
were formed after the enrichment of Bi in the alloys due to the Pb depletion. The oxidation rates of the alloys were much larger than that of the steels, and became larger with higher Pb concentration in the alloys. The compatibility of Pb-Bi alloys with stainless steel was worse when the Pb concentration in the alloys became low, since the dissolution type corrosion was promoted by the Bi composition in the alloy. The Pb-Li alloys were oxidized as they formed LiPbO and LiCO. Then, Li was depleted from the alloy.
Takeda, Toshikazu*; Yokoyama, Kenji; Sugino, Kazuteru
Annals of Nuclear Energy, 109, p.698 - 704, 2017/11
Times Cited Count：2 Percentile：19.37(Nuclear Science & Technology)A new cross section adjustment method has been derived in which systematic errors in measured data and calculated results of neutronics characteristics are estimated and removed in the adjustment. Bias factors which are the ratio between measured data and calculated results are used to estimate systematic errors. The difference of the bias factors from unity is caused generally by systematic errors and stochastic errors. Therefore by determining whether the difference is within the total stochastic errors of measurements and calculations, systematic errors are estimated. Since stochastic errors are determined for individual confidence levels, systematic errors are also dependent to the confidence levels. The method has been applied to cross section adjustments using 589 measured data obtained from fast critical assemblies and fast reactors. The adjustments results are compared with those of the conventional adjustment method. Also the effect of the confidence level to the adjusted cross sections is discussed.
Ono, Ayako; Kurihara, Akikazu; Tanaka, Masaaki; Ohshima, Hiroyuki; Kamide, Hideki; Miyake, Yasuhiro*; Ito, Masami*; Nakane, Shigeru*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
The water experiment apparatus simulating the thermal hydraulics in a reactor vessel under operating the decay heat removal systems (DHRSs) was fabricated. The theoretical evaluation for similarity and results of basic experiments show applicability for a scale model experiment of a sodium-cooled fast reactor. This paper, moreover, describes the results of flow visualization experiment under operating a dipped-type passive DHX, which is planned to be installed in both a loop type reactor and pool type reactor, and the calculation results using FLUENT comparing with the result of water experiment.
Iijima, Susumu*; Kato, Yuichi*; Takasaki, Kenichi*; Okajima, Shigeaki
JAERI-Data/Code 2004-016, 91 Pages, 2004/12
JAERI-Data-Code-2004-016.pdf:7.45MB
The calculation code system "EXPARAM" was designed to analyze the experimental results systematically measured at the fast critical assembly (FCA). Some calculation codes developed independently in JAERI and in US research institutes were collected and arranged as the fast reactor physics calculation code system. The multi-group core calculation code and the perturbation calculation code based on the diffusion theory and the transport theory calculate the reactor physics parameters such as eigenvalue, reaction rate, Doppler reactivity worth and sodium void worth. The dynamic physics parameters such as prompt neutron lifetime and effective delayed neutron fraction are also calculated. Input and Output data of calculation codes are transferred to each other using a direct access file on UNIX computer system.
Iwamura, Takamichi; Okubo, Tsutomu; Akie, Hiroshi; Kugo, Teruhiko; Yonomoto, Taisuke; Kureta, Masatoshi; Ishikawa, Nobuyuki; Nagaya, Yasunobu; Araya, Fumimasa; Okajima, Shigeaki; et al.
JAERI-Research 2004-008, 383 Pages, 2004/06
JAERI-Research-2004-008.pdf:21.49MB
The present report contains the achievement of "Research and Development on Reduced-Moderation Light Water Reactor with Passive Safety Features", which was performed by Japan Atomic Energy Research Institute (JAERI), Hitachi Ltd., Japan Atomic Power Company and Tokyo Institute of Technology in FY2000-2002 as the innovative and viable nuclear energy technology (IVNET) development project operated by the Institute of Applied Energy (IAE). In the present project, the reduced-moderation water reactor (RMWR) has been developed to ensure sustainable energy supply and to solve the recent problems of nuclear power and nuclear fuel cycle, such as economical competitiveness, effective use of plutonium and reduction of spent fuel storage. The RMWR can attain the favorable characteristics such as high burnup, long operation cycle, multiple recycling of plutonium (Pu) and effective utilization of uranium resources based on accumulated LWR technologies.
Okajima, Shigeaki; Yamane, Yoshihiro*; Takemoto, Yoshinari*; Sakurai, Takeshi
Journal of Nuclear Science and Technology, 37(8), p.720 - 723, 2000/08
no abstracts in English
Ogawa, Hironobu; Mukaiyama, Takehiko
JAERI-Tech 99-041, 188 Pages, 1999/05
no abstracts in English
Mukaiyama, Takehiko; Ogawa, Hironobu; *
JAERI-Research 98-001, 76 Pages, 1998/01
JAERI-Research-98-001.pdf:2.97MB
no abstracts in English
W.S.Charlton*; T.A.Parish*; S.Raman*; Shinohara, Nobuo; Ando, Masaki
PHYSOR 96: Int. Conf. on the Physics of Reactors, 3, p.F11 - F20, 1996/00
no abstracts in English
Seimitsu Kikai, 50(9), p.1356 - 1362, 1984/00
no abstracts in English
Nakagawa, Masayuki; *; *
JAERI-M 83-066, 74 Pages, 1983/04
no abstracts in English
; ; ; M.Cho*
JAERI-M 6496, 21 Pages, 1976/03
no abstracts in English
; ; ; M.Cho*
JAERI-M 6067, 21 Pages, 1975/03
no abstracts in English
; *
Journal of Nuclear Science and Technology, 10(3), p.186 - 191, 1973/03
no abstracts in English
; *
Journal of Nuclear Science and Technology, 9(1), p.36 - 46, 1972/01
no abstracts in English
Katsuragi, Satoru;
ENEA CPL News Letter, (13), p.29 - 40, 1972/00
no abstracts in English
Pu fueled system from the data of composite Pu-
U fueled fast critical assemblyNihon Genshiryoku Gakkai-Shi, 4(11), p.746 - 753, 1962/00
no abstracts in English
Okamoto, Koji
no journal, ,
After Fukushima-Daiichi NPP Accidents, the nuclear power plant should be safety first. No more severe accidents should happen around the world. Nuclear energy is and will be a superior resource for sustainable future. To satisfy these requirements, Safety First Nuclear System with advanced technology are strongly expected. The concept of Safety-First Nuclear Systems would be a promising concept after Fukushima. JAEA will continue to develop the Safety-first Nuclear Systems. Also, for the sustainable future, Safety Decommissioning of Fukushima is JAEA's responsibility too.
