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Kubo, Shigenobu; Shimakawa, Yoshio*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06
This paper describes safety design concept for future sodium-cooled fast reactors (SFRs) in Japan, which is based on the safety design criteria and safety design guidelines developed in the auspices of the international forum of generation IV nuclear energy systems. Inherent and/or passive design features are utilized based on SFRs characteristics such as low pressure, high thermal inertia of the system. Lessons learned from the Fukushima Dai-ichi accident is one of important issues to be incorporated into the safety design concept. In order to realize commercial SFRs in the future, robust and rational safety design should be pursued by integrating various factors into the design and limiting additional specific systems, structures and components. Existing engineering principle for the design and manufacturing of SFR's components, and innovative technologies introduced in the FaCT project are keys to achieve the safety concept.
Kubota, Ryuzaburo; Koyama, Kazuya*; Moriwaki, Hiroyuki*; Yamada, Yumi*; Shimakawa, Yoshio*; Suzuki, Toru; Kawada, Kenichi; Kubo, Shigenobu; Yamano, Hidemasa
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
This paper describes an analysis study on the initiating phase of the ATWS events with SAS4A in order to confirm the appropriateness of the core design for the medium-scale SFR (750MWe-1765MWt). Not using a conventional lumping method that multiple fuel sub-assemblies having a similar characteristic were assigned to one channel (representing fuel assembly in SAS4A), each channel represents only the sub-assemblies of identical operating condition. In addition, the detailed power and reactivity distribution were set reflecting the change of insertion position of control rods. Applying these detailed analysis conditions, the SAS4A analyses were performed for unprotected loss-of-flow (ULOF) and unprotected transient overpower (UTOP) during both of the nominal power and the partial power operation. As a result, more proper event progression including incoherency of events especially fuel dispersion after fuel failure was successfully evaluated and then this analysis study suggested that the power excursion with prompt criticality leading to large mechanical energy release can be prevented in the initiating phase of the current design.
Chikazawa, Yoshitaka; Kubo, Shigenobu; Shimakawa, Yoshio*; Kaneko, Fumiaki*; Shoji, Takashi*; Nakata, Shuhei*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
Sodium-cooled reactor (SFR) has superior characteristics thanks to sodium coolant features such as low pressure and high natural convection capability. Involving lessons learned from the 1F accident, requirements on design base DHRS have been modified. In that modification, safety requirements on design extended conditions have been clarified and sodium temperature criteria have been changed taking into account design margin even for design extended conditions. With the new DHRS configuration including ACS, designs of component cooling water system and emergency power supply have been updated.
Chikazawa, Yoshitaka; Kato, Atsushi*; Yamamoto, Tomohiko; Kubo, Shigenobu; Ohno, Shuji; Iwasaki, Mikinori*; Hara, Hiroyuki*; Shimakawa, Yoshio*; Sakaba, Hiroshi*
Nuclear Technology, 196(1), p.61 - 73, 2016/10
Times Cited Count：1 Percentile：10.71(Nuclear Science & Technology)JSFR adopts double boundary for all sodium components. However, design measures are investigated for the secondary sodium fire inside the reactor building, which might be assumed as design extension conditions (DECs). Candidates of sodium fire measures in the secondary sodium systems such as sodium drain, nitrogen injection, pressure release valve, catch pan and leak sodium drain system have been compared from the view point of safety. Wide range of sodium fires in the steam generator room and air cooler have been analyzed evaluating performances of the candidate sodium fire measures.
Chikazawa, Yoshitaka; Kato, Atsushi; Hayafune, Hiroki; Shimakawa, Yoshio*; Kamishima, Yoshio*
Nuclear Technology, 192(2), p.111 - 124, 2015/11
Times Cited Count：1 Percentile：9.74(Nuclear Science & Technology)Evaluation of severe external hazards on JSFR has been analyzed. For seismic design, safety components are confirmed to maintain their functions even against recent strong earthquakes. For tsunam, hypothetical station blackout has been evaluated.
Chikazawa, Yoshitaka; Kato, Atsushi; Yamamoto, Tomohiko; Kubo, Shigenobu; Iwasaki, Mikinori*; Hara, Hiroyuki*; Shimakawa, Yoshio*; Sakaba, Hiroshi*
Proceedings of 2014 International Congress on the Advances in Nuclear Power Plants (ICAPP 2014) (CD-ROM), p.523 - 530, 2014/04
JSFR adopts double boundary for all sodium components. However, design measures are investigated for the secondary sodium fire inside the reactor building, which might be assumed as design extension conditions (DECs). Candidates of sodium fire measures in the secondary sodium systems such as sodium drain, nitrogen injection, pressure release valve, catch pan and leak sodium drain system have been compared from the view point of safety. Wide range of sodium fires in the steam generator room have been analyzed evaluating performances of the candidate sodium fire measures.
Kato, Atsushi; Kubo, Shigenobu; Chikazawa, Yoshitaka; Hayafune, Hiroki; Yokoi, Shinobu*; Nakata, Shuhei*; Tani, Akihiro*; Shimakawa, Yoshio*
Proceedings of 2014 International Congress on the Advances in Nuclear Power Plants (ICAPP 2014) (CD-ROM), p.616 - 623, 2014/04
This paper focuses on loss of heat removal system (LOHRS) type event as Design Extension Condition (DEC) and describes candidates design measures to improve the decay heat removal system of JSFR against LOHRS type DEC. The design requirements are determined based on the Safety Design Criteria for Generation-IV Sodium-cooled fast reactor system. Effectiveness and reliability of the candidate design measures are discussed with preliminary evaluations.
Hayafune, Hiroki; Kato, Atsushi; Chikazawa, Yoshitaka; Okubo, Tsutomu; Sagawa, Hiroshi*; Shimakawa, Yoshio*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 11 Pages, 2013/03
Evaluation of earthquake and tsunami on JSFR has been analyzed. For seismic design, safety components are confirmed to maintain their functions even against recent strong earthquakes. As for Tsunami, some parts of reactor building might be submerged including component cooling water system whose final heat sink is sea water. However, in the JSFR design, safety grade components are independent from component cooling water system (CCWS). The JSFR emergency power supply adopts a gas turbine system with air cooling, since JSFR does not basically require quick start-up of the emergency power supply thanks to the natural convection DHRS. Even in case of long station blackout, the DHRS could be activated by emergency batteries or manually and be operated continuously by natural convection.
Kubo, Shigenobu; Yamano, Hidemasa; Chikazawa, Yoshitaka; Shimakawa, Yoshio*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 9 Pages, 2013/03
This paper describes the safety design approach for JSFR. To achieve safety goals for Generation IV reactor, design measures should be enhanced against design extension conditions including those for external events considering the lessons learned from the TEPCO's Fukushima Dai-ichi Nuclear Power Plants accident. The current safety design approach for JSFR intends to meet the safety design criteria for Generation-IV SFR developed in the framework of the Generation-IV International Forum. Design extension conditions and related design measures are identified and selected with due consideration of the safety features of SFR. Design approach and measures for severe external events such as earthquake and tsunami, external missiles, failure to shutdown type events and failure to heat removal type events are shown. Several situations to be practically eliminated are proposed with possible design measures.
Kubota, Ryuzaburo*; Yamada, Yumi*; Koyama, Kazuya*; Shimakawa, Yoshio*; Yamano, Hidemasa; Kubo, Shigenobu; Suzuki, Toru; Tobita, Yoshiharu
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
This paper describes a melt-down event progression revealed by a numerical simulation in the protected loss of heat sink (PLOHS) event for Japan Sodium-cooled Fast Reactor (JSFR). A multi-component multi-field computer code, MUTRAN, has been applied in order to simulate complicated core material motions and associated heat-transfer phenomena among the materials in a degradation core. The analyses with MUTRAN covered core degradation behaviors from the intact geometry and addressed the two initial states: one was the core without the coolant as the leakage type, and the other was the core covered by the coolant only up to the top of the fissile fuel as the boiling type. The analyses revealed representative event progression.
Chikazawa, Yoshitaka; Enuma, Yasuhiro; Kisohara, Naoyuki; Yamano, Hidemasa; Kubo, Shigenobu; Hayafune, Hiroki; Sagawa, Hiroshi*; Okamura, Shigeki*; Shimakawa, Yoshio*
Proceedings of 2012 International Congress on Advances in Nuclear Power Plants (ICAPP '12) (CD-ROM), p.677 - 686, 2012/06
Evaluation of Earthquake and Tsunami on JSFR has been analyzed. For seismic design, safety components are confirmed to maintain their functions even against recent strong Earthquakes. As for Tsunami, some parts of reactor building might be submerged including component cooling water system whose final heat sink is sea water. However, in the JSFR design, safety grade components are independent from component cooling water system (CCWS). The JSFR emergency power supply adopts a gas turbine system with air cooling, since JSFR does not basically require quick start-up of the emergency power supply thanks to the natural convection DHRS. Even in case of long station blackout, the DHRS could be activated by emergency batteries or manually and be operated continuously by natural convection.
Kubo, Shigenobu*; Shimakawa, Yoshio*; Yamano, Hidemasa; Kotake, Shoji
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00
In order to embody a safety design, a higher level safety principle was broken down into a set of design requirements for each safety related system, structure and component (SSC). This paper will present an output of the safety requirements for safety related SSCs of JSFR.
Yamano, Hidemasa; Kubo, Shigenobu; Shimakawa, Yoshio*; Fujita, Kaoru; Suzuki, Toru; Kurisaka, Kenichi
Science and Technology of Nuclear Installations, 2012, p.614973_1 - 614973_14, 2012/00
Times Cited Count：14 Percentile：71.22(Nuclear Science & Technology)This paper describes safety requirements for JSFR conformed to the defense-in-depth principle in IAEA. The safety design accommodation in JSFR was validated by safety analyses for representative DBEs: primary pump seizure and long-term loss-of-offsite power accidents. The safety analysis also showed the effectiveness of the passive shutdown system and mitigation measures against a typical ATWS.
Tani, Akihiro*; Shimakawa, Yoshio*; Kubo, Shigenobu*; Fujimura, Ken; Yamano, Hidemasa
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10
Yamano, Hidemasa; Kubo, Shigenobu*; Shimakawa, Yoshio*; Fujita, Kaoru; Suzuki, Toru; Kurisaka, Kenichi
Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.728 - 740, 2011/05
Kubo, Shigenobu*; Shimakawa, Yoshio*; Yamano, Hidemasa; Kotake, Shoji
Journal of Nuclear Science and Technology, 48(4), p.547 - 555, 2011/04
Safety design requirements for JSFR were summarized taking the development targets of FaCT project and design feature of JSFR into account. The related safety principle and requirements for Monju, CRBRP, PRISM, SPX, LWRs, IAEA standards, goals of GIF and basic principle of INPRO etc. were also taken into account so that the safety design requirements can be a next-generation global-standard. The development targets for safety and reliability are set based on that of FaCT. Namely, ensuring safety and reliability equal to future LWR and related fuel cycle facilities. In order to achieve these targets, the defence-in-depth philosophy is used as the basic safety design principle. General features of the safety design requirements are (1) Achievement of higher reliability, (2) Achievement of higher inspectability and maintainability, (3) Introduction of passive safety features, (4) Reduction of operator action needs, (5) Design consideration against Beyond Design Basis Events, (6) In Vessel Retention of degraded core materials, (7) Prevention and mitigation against sodium chemical reactions, (8) Design against external events. Current specific requirements for the each system and component are summarized taking the basic design concept of JSFR into account, which is an advanced loop type large output power plant with mixed oxide fuelled core.
Kurisaka, Kenichi; Shimakawa, Yoshio*
Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.1220 - 1227, 2010/06
JAEA participates in activities of the Generation IV International Forum's Risk and Safety Working Group (GIF/RSWG). The GIF/RSWG has developed the Integrated Safety Assessment Methodology (ISAM), which consists of five distinct analytical tools; i.e., Qualitative Safety Features Review, Phenomena Identification and Ranking Table (PIRT), Objective Provision Tree (OPT), Deterministic and Phenomenological Analyses (DPA) and Probabilistic Safety Analysis (PSA). Among them, PIRT, OPT, DPA and PSA were applied preliminarily to the Japanese Sodium-cooled Fast Reactor (JSFR) system. The JSFR system includes major innovative safety features such as Self-Actuated Shutdown System (SASS), passive decay heat removal system. PIRT was applied to examination of the reactor safe shutdown by means of SASS during a loss-of-flow accident with a failure of conventional reactor shutdown system. OPTs were developed to assess the structure of safety architecture of the JSFR in an adequate manner based on the defence-in-depth philosophy. Some provisions explicitly shown in the OPT are characterized with the safety design requirements of decay heat removal function. Sufficiency to those requirements was confirmed by DPA. PSA was conducted with analytical models, which were based on those OPT and DPA results. The PSA served to quantification of the level of safety and to the system design improvement in the JSFR.
Yamano, Hidemasa; Kubo, Shigenobu*; Kurisaka, Kenichi; Shimakawa, Yoshio*; Sago, Hiromi*
Nuclear Technology, 170(1), p.159 - 169, 2010/04
Times Cited Count：15 Percentile：70.35(Nuclear Science & Technology)An advanced large-scale sodium-cooled fast reactor (named JSFR) adopts an innovative two-loop cooling system. This cooling system design gives rise to major technological issues: hydraulic and structural integrity due to the increase in one-loop coolant flow rate, safety design against the break or failure in one-loop piping and ensuring the reliability of decay heat removal system. The present paper describes that the piping structural integrity due to flow-induced vibration has been investigated using a 1/3-scale hot-leg piping test. The structural integrity of the hot-leg piping in the JSFR design has been confirmed by a flow-induced-vibration analytical methodology, verified with the experimental data. Additional experimental results have revealed that hydraulic issues including gas entrainment and vortex cavitation could be prevented by some design measures. By applying appropriate safety design, the two-loop system has been confirmed to be valid against the break or failure in one-loop piping by a safety evaluation in this study. The decay heat removal system with natural circulation is designed in conformity with the two-loop system by introducing adequate safety designs. In this paper, the validity of this decay heat removal system is given by a probabilistic safety assessment and safety evaluation.
Nakanishi, Shigeyuki*; Hosoya, Takusaburo; Kubo, Shigenobu*; Kotake, Shoji; Takamatsu, Misao; Aoyama, Takafumi; Ikarimoto, Iwao*; Kato, Jungo*; Shimakawa, Yoshio*; Harada, Kiyoshi*
Nuclear Technology, 170(1), p.181 - 188, 2010/04
Times Cited Count：14 Percentile：67.77(Nuclear Science & Technology)A self-actuated shutdown system (SASS) for sodium cooled fast reactor (SFR) is a passive safety feature which inserts control rods by the gravity force, where the detachment of the rods would be achieved by the coolant temperature rise under anticipated transient without scram (ATWS) conditions. Various out-of-pile tests have already carried out to investigate the basic characteristics of SASS, and a demonstration test of holding stability under the reactor operation condition has been performed, where a function test of the driving system to re-connect and of pulling out the control rod have been done in the experimental reactor JOYO. The element irradiation tests have been also conducted to confirm that no impact will be foreseen by the irradiation. The effectiveness of SASS for a reference core design of JSFR has been evaluated through all types of ATWS. As a result, it is ensured that JSFR will have a reliable passive shutdown system.
Yamano, Hidemasa; Kubo, Shigenobu; Kurisaka, Kenichi; Shimakawa, Yoshio*; Sago, Hiromi*
Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.469 - 504, 2008/06
The conceptual design of an advanced sodium-cooled fast reactor (named JSFR) is currently carried out by the Japan Atomic Energy Agency (JAEA). In general, a large-scale nuclear reactor (approximately 1.5 GWe class) tended to increase in the number of loops (e.g., four loops in Super Ph
nix and APWR), while the JSFR adopts a two-loop cooling system that enables significantly reducing a plant construction cost resulting from decreasing in material amount of the nuclear steam supply system and in the reactor building volume. This paper describes technological feasibility of the two-loop cooling system in JSFR; especially, focused on flow-induced vibration of piping, safety analysis and decay heat removal system.
