Kubo, Shigenobu; Chikazawa, Yoshitaka; Ohshima, Hiroyuki; Uchida, Masato*; Miyagawa, Takayuki*; Eto, Masao*; Suzuno, Tetsuji*; Matoba, Ichiyo*; Endo, Junji*; Watanabe, Osamu*; et al.
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
The authors are developing the design concept of pool-type sodium-cooled fast reactor (SFR) that addresses Japan's specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDGs) for Generation IV SFRs. The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. A design concept of 1,500 MWt (650 MWe) class pool-type SFR was thought up by applying design technology obtained from the design of advanced loop-type SFR, named JSFR, equipped with safety measures that reflect results from the feasibility study on commercialized fast reactor cycle systems and fast reactor cycle technology development, improved maintainability and repairability, and lessons learned from the Fukushima Daiichi Nuclear Power Plants accident.
Partitioning and Transmutation Technology Division, Nuclear Science and Engineering Center
JAEA-Technology 2017-033, 383 Pages, 2018/02
JAEA is pursuing research and development (R&D) on volume reduction and mitigation of degree of harmfulness of high-level radioactive waste. Construction of Transmutation Experimental Facility (TEF) is under planning as one of the second phase facilities in the Japan Proton Accelerator Complex (J-PARC) program to promote R&D on the transmutation technology with using accelerator driven systems (ADS). The TEF consists of two facilities: ADS Target Test Facility (TEF-T) and Transmutation Physics Experimental Facility (TEF-P). Development of spallation target technology and study on target materials are to be conducted in TEF-T with impinging a high intensity proton beam on a liquid lead-bismuth eutectic target. Whereas in TEF-P, by introducing a proton beam to minor actinide loaded cores, reactor physical properties of the cores are to be studied, and operation experiences of ADS are to be acquired. This report summarizes results of safety design for establishment permit of one of two TEF facilities, TEF-P.
Nakagawa, Shigeaki; Sato, Hiroyuki; Fukaya, Yuji; Tokuhara, Kazumi; Ohashi, Hirofumi
JAEA-Technology 2017-022, 32 Pages, 2017/09
As for the design of commercial HTGRs, the fuel design, core design, reactor coolant system design, secondary helium system design, decay heat removal system design and confinement system design are very important and quite different from those of LWRs. To contribute the establishment of the safety standards for commercial HTGRs, the evaluation items to attain safety requirements in fuel and core designs were studied. In this study, the excellence features of HTGRs based on passive safety or inherent safety were fully reflected. Additionally, concerning the core design, the stability to spatial power oscillation in reactor core of HTGR was studied. The evaluation items as the result of the study are applicable to the safety design of commercial HTGRs in the future.
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
The GIF Safety Design Criteria Task Force (SDC TF) has been developing a set of safety design guidelines (SDG) to support practical application of SDC since the completion of the "SDC Phase I Report" that clarifies safety design requirements for Gen-IV SFR systems. The main objective of the SDG development is to assist SFR developers and vendors to utilize the SDC in their design process for improving the safety in specific topical areas including the use of inherent/passive safety features and the design measures for prevention and mitigation of severe accidents. The first report on "Safety Approach SDGs" aims to provide guidance on safety approaches covering specific safety issues on fast reactor core reactivity and on loss of heat removal. The second report on "SDGs on key Structures, Systems and Components (SSCs)" focuses on the functional requirements for SSCs important to safety; reactor core system, reactor coolant system, and containment system.
Chikazawa, Yoshitaka; Kato, Atsushi; Nabeshima, Kunihiko; Otaka, Masahiko; Uzawa, Masayuki*; Ikari, Risako*; Iwasaki, Mikinori*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05
Design study and evaluation for SDC and safety SDG on the BOP of the demonstration JSFR including fuel handling system, power supply system, component cooling water system, building arrangement are reported. For the fuel handling system, enhancement of storage cooling system has been investigated adding diversified cooling systems. For the power supply, existing emergency power supply system has been reinforced and alternative emergency power supply system is added. For the component cooling system and air conditioning, requirements and relation between safety grade components are investigated. Additionally for the component cooling system, design impact when adding decay heat removal system by sea water has been investigated. For reactor building, over view of evaluation on the external events and design policy for distributed arrangement is reported. Those design study and evaluation provides background information of SDC and SDG.
Osaki, Hirotaka; Shimazaki, Yosuke; Sumita, Junya; Shibata, Taiju; Konishi, Takashi; Ishihara, Masahiro
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05
For the design on the VHTR graphite components, it is desirable to employ graphite material with higher strength. IG-430 graphite has been developed as an advanced candidate for VHTR. However, the new developed IG-430 does not have enough databases for the design of HTGR. In this paper, the compressive strength (Cs) of IG-430, one of important strengths for design data, is statistically evaluated. The component reliability is evaluated based on the safety factors defined by the graphite design code, and the applicability as the VHTR graphite material is discussed. It was found that IG-430 has higher strength (about 11%) and lower standard deviation (about 27%) than IG-110 which is one of traditional graphites used for HTGR, because the crack in IG-430 would not easy to propagate rather than IG-110. Since fracture probability for IG-430 is low, the higher reliability of core-component will be achieved using IG-430. It is expected that IG-430 is applicable for VHTR graphite material.
Nakashima, Hiroshi; Nakane, Yoshihiro; Masukawa, Fumihiro; Matsuda, Norihiro; Oguri, Tomomi*; Nakano, Hideo*; Sasamoto, Nobuo*; Shibata, Tokushi*; Suzuki, Takenori*; Miura, Taichi*; et al.
Radiation Protection Dosimetry, 115(1-4), p.564 - 568, 2005/12
The High Intensity Proton Accelerator Project, named as J-PARC, is in progress, aiming at studies on the latest basic science and the advancing nuclear technology. In the project, the high-energy proton accelerator complex of the world highest intensity is under construction. In order to establish a reasonable shielding design, both simplified and detailed design methods were used in the shielding design of J-PARC. This paper reviews the present status of the radiation safety design study for J-PARC.
Takada, Eiji*; Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Shimakawa, Satoshi; Nojiri, Naoki; Fujimoto, Nozomu
JAERI-Tech 2004-048, 60 Pages, 2004/06
The HTTR (High Temperature Engineering Test Reactor), which has thermal output of 30MW, coolant inlet temperature of 395C and coolant outlet temperature of 850C/950C, is a first high temperature gas-cooled reactor (HTGR) in Japan. The HTGR has a high inherent safety potential to accident condition. Safety demonstration tests using the HTTR are underway in order to demonstrate such excellent inherent safety features of the HTGR. The reactivity insertion test demonstrates that rapid increase of reactor power by withdrawing the control rod is restrained by only the negative reactivity feedback effect without operating the reactor power control system, and the temperature transient of the reactor is slow. The best estimated analyses have been conducted to simulate reactor transients during the reactivity insertion test. A one-point core dynamics approximation with one fuel channel model is applied to this analysis. It was found that the analytical model for core dynamics could simulate the reactor power behavior.
Sakaba, Nariaki; Nakagawa, Shigeaki; Furusawa, Takayuki; Tachibana, Yukio
JAERI-Tech 2004-045, 67 Pages, 2004/04
Safety demonstration tests using the HTTR are now underway in order to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to research and development for the VHTR, which is one of the Generation IV reactors. In the safety demonstration tests, the coolant flow reduction test by tripping one or two out of three gas circulators is being performed between FY2002 and FY 2005 and by tripping all the three gas circulators will be conducted after FY2006. This paper describes the structural integrity assessment of the primary pressurised water cooler after one and two gas circulators run down. Also, the possibility of natural convection in the primary coolant after all the three gas circulator stopped was evaluated by the operation data of the reactor-scram test performed during the rise-to-power tests.
Sakaba, Nariaki; Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Takada, Eiji*; Saito, Kenji; Furusawa, Takayuki; Tochio, Daisuke; Tachibana, Yukio; Iyoku, Tatsuo
JAERI-Tech 2004-014, 24 Pages, 2004/02
Safety demonstration tests using the HTTR are in progress to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to not only the commercial HTGRs but also the research and development for the VHTR one of the Generation IV reactors. This paper describes the reactivity insertion test and coolant flow reduction test by trip of gas circulator and partial flow loss of coolant planned in 2004 with detailed test method, procedure and results of pre-test analysis. From the analytical results, it was found that the negative reactivity feedback effect of the core brings the reactor power safely to a stable level without a reactor scram.
Nabeshima, Kunihiko; Nakatsuka, Toru; Ishikawa, Nobuyuki; Uchikawa, Sadao
JAERI-Conf 2003-020, 240 Pages, 2003/11
The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since 1998 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The workshop began with five lectures on status of research and development on RMWRs in JAERI entitled "Status and Future Program of Research and Development on Reduced-Moderation Water Reactors", "Design of Small Reduced-Moderation Water Reactors", "Critical Experiments for Reduced-Moderation Water Reactors", "Critical Heat Flux Experiments in Tight Lattice Core" and "Development of High Performance Cladding". Then two lectures followed: "Status of Phase II of Feasibility Studies on Commercialized Fast Breeder Reactor System" by JNC and "Present Status of Study on Super-critical water Cooled Power Reactor" by Toshiba Corporation.
Sawa, Kazuhiro; Ueta, Shohei; Iyoku, Tatsuo
Proceedings of International Conference on Global Environment and Advanced Nuclear Power Plants (GENES4/ANP 2003) (CD-ROM), 10 Pages, 2003/09
This paper provides present status of research and development for the coated fuel particle (CFPs) including the advanced ZrC-CFP. Current HTGR employs so-called TRISO-CFPs with SiC layer. In safety design of the HTGR fuels, it is important to retain fission products within CFPs so that their release to primary coolant does not exceed an acceptable level. The behavior of TRISO-CFPs has been investigated through experiments and reactor operation. These data show excellent performance of the TRISO-CFPs when they are correctly fabricated. On the other hand, the crystalline material comprising the SiC layer has a tendency to decompose at high temperature. The transition temperatures of beta-SiC (as-deposited) to alpha-SiC vary from 1600 to 2200C. ZrC is one of the transition metal carbides which are characterized by the high melting point and the thermodynamic stability etc. The CFPs with CVD-ZrC coatings have been investigated including the fabrication processes and characterization techniques developments.
Sakaba, Nariaki; Nakagawa, Shigeaki; Takada, Eiji*; Tachibana, Yukio; Saito, Kenji; Furusawa, Takayuki; Takamatsu, Kuniyoshi; Tochio, Daisuke; Iyoku, Tatsuo
JAERI-Tech 2003-074, 37 Pages, 2003/08
Safety demonstration tests using HTTR are now underway in order to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to research and development for the VHTR, which is one of the Generation IV reactors. The first phase of the safety demonstration tests includes reactivity insertion tests by means of control-rod withdrawal and coolant flow reduction tests by tripping the gas circulators. In the second phase, accident simulation tests will be conducted. This paper describes the plan of coolant flow reduction tests by tripping of gas circulators planned in August 2003 with detailed test method, procedure and results of pre-test analysis. The analysis results of the steady state and transient behaviours of the reactor and the plant of the HTTR show that in the case of a rapid decrease of the coolant flow rate, the negative reactivity feedback effect of the core brings the reactor power safely to certain stable level without a reactor scram, and that the temperature transient of the reactor core is slow.
Nakagawa, Shigeaki; Sakaba, Nariaki; Takada, Eiji*; Tachibana, Yukio; Saito, Kenji; Furusawa, Takayuki; Sawa, Kazuhiro
JAERI-Tech 2003-049, 22 Pages, 2003/03
Safety demonstration tests in the HTTR (High Temperature Engineering Test Reactor) will be carried out in order to verify inherent safety features of the HTGR (High Temperature Gas-cooled Reactor). The first phase of the safety demonstration tests includes the reactivity insertion test by the control rod withdrawal and the coolant flow reduction test by the gas circulator trip. In the second phase, accident simulation tests will be conducted. By comparison of their experimental and analytical results, the prediction capability of the safety evaluation codes such as the core and the plant dynamics codes will be improved and verified, which will contribute to establish the safety design and the safety evaluation technologies of the HTGRs. The results obtained through its safety demonstration tests will be also utilised for the establishment of the safety design guideline, the safety evaluation guideline, etc. This paper describes the test program of the overall safety demonstration tests and the test method, the test conditions and the results of the pre-test analysis of the reactivity insertion test and the partial gas circulator trip test planned in March 2003.
Nishihara, Tetsuo; Shimizu, Akira; Tanihira, Masanori*; Uchida, Shoji*
JAERI-Tech 2002-101, 46 Pages, 2003/01
no abstracts in English
Sakaba, Nariaki; Nakagawa, Shigeaki; Tachibana, Yukio
Proceedings of GLOBAL2003 Atoms for Prosperity; Updating Eisenhower's Global Vision for Nuclear Energy (CD-ROM), p.293 - 299, 2003/00
Safety demonstration tests using the HTTR are now underway in order to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to research and development for the VHTR, which is one of the Generation IV reactors. The first phase of the safety demonstration tests includes the reactivity insertion test by means of control-rod withdrawal and the coolant flow reduction test by tripping the gas circulators. The coolant flow reduction tests are simulation tests of anticipated transients without scram (ATWS). In the second phase of the safety demonstration tests, accident simulation tests will be conducted. This paper describes the plan of the overall safety demonstration tests and coolant flow reduction tests with test method, test conditions, and analytical and experimental results. From the results, it was found that the negative reactivity feedback of the core brings the reactor power safely to a stable level without a reactor scram in the case of a rapid decrease of the coolant flow rate after tripping of gas circulators.
Tada, Eisuke; Hada, Kazuhiko; Maruo, Takeshi; Safety Design/Evaluation Group
Purazuma, Kaku Yugo Gakkai-Shi, 78(11), p.1145 - 1156, 2002/11
no abstracts in English
Nishihara, Tetsuo; Muto, Yasushi; Uchida, Shoji*; Yoshioka, Naoki*
JAERI-Tech 2001-077, 44 Pages, 2001/12
JAERI has conducted the feasibility study of the HTGR gas turbine system from 1996 to 2000 sponsored by MEXT. This report concludes the safety criteria and rationalization of the safety items in the HTGR system.With respect to the safety criteria, the same value for the LWR is selected as the limit of radiation exposure. Probability of the design basis event (DBE) and beyond design basis event (BDBE) is set lower than those for the LWR to get higher safety margin. Adequate initial events and mitigation system are selected to consider the event sequence. The concept of the probability analysis is applied to identify DBEs and BDBEs. It is found that some safety items can be rationalized in consideration of the safety features of the HTGR. Finally, the safety class and design category of the items in the HGTR-GT are classified.
Nakagawa, Shigeaki; Saikusa, Akio; Kunitomi, Kazuhiko
Nuclear Technology, 133(2), p.141 - 152, 2001/02
no abstracts in English
Nishihara, Tetsuo; Hada, Kazuhiko
Nippon Genshiryoku Gakkai-Shi, 41(5), p.571 - 578, 1999/05
no abstracts in English