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Journal Articles

The Development status of Generation IV reactor systems, 2; High temperature gas-cooled reactor (HTGR)

Kunitomi, Kazuhiko; Nishihara, Tetsuo; Yan, X.; Tachibana, Yukio; Shibata, Taiju

Nippon Genshiryoku Gakkai-Shi, 60(4), p.236 - 240, 2018/04

High temperature gas-cooled reactor (HTGR) is a graphite-moderated and helium-gas-cooled thermal-neutron reactor that has excellent safety features and can produce high temperature heat of 950$$^{circ}$$C. It is expected to use for various heat applications as well as for electricity generation to reduce carbon dioxide emission. Japan Atomic Energy Agency (JAEA) has been promoted research and development to demonstrate the HTGR safety features using High temperature engineering test reactor (HTTR) and it's heat application. JAEA are also conducting the action to international deployment of Japanese HTGR technologies in cooperation with industries-government-academia. This paper reports status of the research and development of HTGR and domestic and international collaborations.

JAEA Reports

Evaluation items to attain safety requirements in fuel and core designs for commercial HTGRs

Nakagawa, Shigeaki; Sato, Hiroyuki; Fukaya, Yuji; Tokuhara, Kazumi; Ohashi, Hirofumi

JAEA-Technology 2017-022, 32 Pages, 2017/09

JAEA-Technology-2017-022.pdf:3.59MB

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.

Journal Articles

Improvement of core dynamics analysis of control rod withdrawal test in HTGR

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki

Nippon Genshiryoku Gakkai Wabun Rombunshi, 5(1), p.45 - 56, 2006/03

The HTTR (High Temperature Engineering Test Reactor), which has thermal output of 30MW, coolant inlet temperature of 395$$^{circ}$$C and coolant outlet temperature of 850$$^{circ}$$C/950$$^{circ}$$C, 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. A one-point core dynamics approximation with one fuel channel model had applied to this analysis. It was found that the analytical model for core dynamics couldn't simulate the reactor power behavior accurately. This report proposes an original method using temperature coefficients of some regions in the core. It is crucial to evaluate this method precisely to simulate a performance of HTGR during the test.

Journal Articles

An Approach for development of technical structural standard in ITER

Nakahira, Masataka; Takeda, Nobukazu

Hozengaku, 4(4), p.47 - 52, 2006/01

The technical structural standard for ITER (International Thermonuclear Experimental Fusion Reactor) should be innovative because of their quite different features of safety and mechanical components from nuclear fission reactors, and the necessity of introducing several new fabrication and examination technologies. Recognizing the international importance of Fusion Standard, Japan and ASME has started the cooperation development of the Fusion Standard. This paper shows the special features of ITER from view points of safety, design and fabrication, and proposes approach for development of the fusion standard.

JAEA Reports

Rationalization and utilization of double-wall vacuum vessel for tokamak fusion facility

Nakahira, Masataka

JAERI-Research 2005-030, 182 Pages, 2005/09

JAERI-Research-2005-030.pdf:12.57MB

It is difficult for Vacuum Vessel (VV) of ITER to apply a non-destructive in-service inspection (ISI) and then new safety concept is needed. Present fabrication standards are not applicable to the VV, because the access is limited to the backside of closure weld of double wall. Fabrication tolerance of VV is $$pm$$5mm even the structure is huge as high as 10m. This accuracy requires a rational method on the estimation of welding deformation. In this report, an inherent safety feature of the tokamak is proved closing up a special characteristic of termination of fusion reaction due to tiny water leak. A rational concept not to require ISI without sacrificing safety is shown based on this result. A partial penetration T-welded joint is proposed to establish a rational fabrication method of double wall. Strength and susceptibility to crevice corrosion is evaluated for this joint and feasibility is confirmed. A rational method of estimation of welding deformation for large and complex structure is proposed and the efficiency is shown by comparing analysis experimental results of full-scale test.

JAEA Reports

Safety demonstration test (SR-3/S1C-3/S2C-3/SF-2) plan using the HTTR (Contract research)

Nakagawa, Shigeaki; Sakaba, Nariaki; Takamatsu, Kuniyoshi; Takada, Eiji*; Tochio, Daisuke; Owada, Hiroyuki*

JAERI-Tech 2005-015, 26 Pages, 2005/03

JAERI-Tech-2005-015.pdf:1.77MB

Safety demonstration tests using the HTTR are in progress since 2002 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 reactor candidates. This paper describes the reactivity insertion test (SR-3), the coolant flow reduction test by tripping of gas circulators (S1C-3/S2C-3), and the partial flow loss of coolant test (SF-2) planned in March 2005 with their 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.

Journal Articles

Achievement of reactor-outlet coolant temperature of 950$$^{circ}$$C in HTTR

Fujikawa, Seigo; Hayashi, Hideyuki; Nakazawa, Toshio; Kawasaki, Kozo; Iyoku, Tatsuo; Nakagawa, Shigeaki; Sakaba, Nariaki

Journal of Nuclear Science and Technology, 41(12), p.1245 - 1254, 2004/12

 Times Cited Count:66 Percentile:2.59(Nuclear Science & Technology)

A High Temperature Gas-cooled Reactor (HTGR) is particularly attractive due to its capability of producing high-temperature helium gas and to its inherent safety characteristics. The High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, achieved its rated thermal power of 30MW and reactor-outlet coolant temperature of 950$$^{circ}$$C on 19 April 2004. During the high-temperature test operation which is the final phase of the rise-to-power tests, reactor characteristics and reactor performance were confirmed, and reactor operations were monitored to demonstrate the safety and stability of operation. The reactor-outlet coolant temperature of 950$$^{circ}$$C makes it possible to extend high-temperature gas-cooled reactor use beyond the field of electric power. Also, highly effective power generation with a high-temperature gas turbine becomes possible, as does hydrogen production from water. The achievement of 950$$^{circ}$$C will be a major contribution to the actualization of producing hydrogen from water using the high-temperature gas-cooled reactors. This report describes the results of the high-temperature test operation of the HTTR.

Journal Articles

Performance test of HTTR

Nakagawa, Shigeaki; Tachibana, Yukio; Takamatsu, Kuniyoshi; Ueta, Shohei; Hanawa, Satoshi

Nuclear Engineering and Design, 233(1-3), p.291 - 300, 2004/10

 Times Cited Count:6 Percentile:59.89(Nuclear Science & Technology)

The High Temperature Gas-cooled Reactor (HTGR) is particularly attractive due to its capability of producing high temperature helium gas and due to its inherent safety characteristics. The High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, was successfully constructed at the Oarai Research Establishment of the Japan Atomic Energy Research Institute. The HTTR achieved full power of 30MW at a reactor outlet coolant temperature of about 850$$^{circ}$$C on December 7, 2001 during the "rise-to-power tests". Two kinds of tests were carried out during the "rise-to-power tests". One is commissioning test to get operation permit by the government and another is test to confirm a performance of the reactor, heat exchanger, control system. From the test results of the "rise-to-power tests" up to 30MW, the functionality of the reactor and the cooling system were confirmed, and it was also confirmed that an operation of reactor facility can be performed safely.

Journal Articles

Safety demonstration tests using high temperature engineering test reactor

Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Tachibana, Yukio; Sakaba, Nariaki; Iyoku, Tatsuo

Nuclear Engineering and Design, 233(1-3), p.301 - 308, 2004/10

 Times Cited Count:19 Percentile:19.84(Nuclear Science & Technology)

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are conducted for demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) as well as for providing core and plant transient data for validation of HTGR safety analysis codes. The safety demonstration tests are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase tests simulating reactivity insertion events and coolant flow reduction events started in FY 2002. The first phase safety demonstration tests will continue until FY 2005, and the second phase tests will be carried out from FY 2006.

JAEA Reports

Core dynamics analysis of control rod withdrawal test in HTTR (Contract Research)

Takada, Eiji*; Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Shimakawa, Satoshi; Nojiri, Naoki; Fujimoto, Nozomu

JAERI-Tech 2004-048, 60 Pages, 2004/06

JAERI-Tech-2004-048.pdf:4.18MB

The HTTR (High Temperature Engineering Test Reactor), which has thermal output of 30MW, coolant inlet temperature of 395$$^{circ}$$C and coolant outlet temperature of 850$$^{circ}$$C/950$$^{circ}$$C, 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.

JAEA Reports

Structural integrity assessment of helium component during safety demonstration test using HTTR, 1 (Contract Research)

Sakaba, Nariaki; Nakagawa, Shigeaki; Furusawa, Takayuki; Tachibana, Yukio

JAERI-Tech 2004-045, 67 Pages, 2004/04

JAERI-Tech-2004-045.pdf:4.74MB

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.

JAEA Reports

Safety demonstration test (SR-2/S2C-2/SF-1) plan using the HTTR (Contract research)

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

JAERI-Tech-2004-014.pdf:1.06MB

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.

JAEA Reports

Safety demonstration test (S1C-2/S2C-1) plan using the HTTR (Contract research)

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

JAERI-Tech-2003-074.pdf:1.83MB

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.

JAEA Reports

Safety demonstration test (SR-1/S1C-1) plan of HTTR (Contract research)

Nakagawa, Shigeaki; Sakaba, Nariaki; Takada, Eiji*; Tachibana, Yukio; Saito, Kenji; Furusawa, Takayuki; Sawa, Kazuhiro

JAERI-Tech 2003-049, 22 Pages, 2003/03

JAERI-Tech-2003-049.pdf:1.17MB

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.

Journal Articles

Safety demonstration test plan of HTTR; Overall program and result of coolant flow reduction test

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.

Journal Articles

Approach for safety assurance and structural integrity of ITER

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

Journal Articles

Development of a simulation model and safety evaluation for depressurization accident without reactor scram in an advanced HTGR

Nakagawa, Shigeaki; Saikusa, Akio; Kunitomi, Kazuhiko

Nuclear Technology, 133(2), p.141 - 152, 2001/02

 Times Cited Count:2 Percentile:77.64(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Method and results of safety evaluation of the high-temperature engineering test reactor

Nakagawa, Shigeaki; Kunitomi, Kazuhiko; Sawa, Kazuhiro

Nuclear Technology, 115(3), p.266 - 280, 1996/09

 Times Cited Count:3 Percentile:64.88(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Passive heat removal by vessel cooling system of HTTR during no forced cooling accidents

Kunitomi, Kazuhiko; Nakagawa, Shigeaki; Shinozaki, Masayuki

Nucl. Eng. Des., 166(2), p.179 - 190, 1996/00

 Times Cited Count:19 Percentile:16

no abstracts in English

Journal Articles

Safety concept of severe accident free HTR and its demonstration by the HTTR

Kunitomi, Kazuhiko; Shiozawa, Shusaku; Baba, Osamu

Proc. of ASME$$cdot$$JSME 4th Int. Conf. on Nuclear Engineering 1996 (ICONE-4), 2, p.303 - 307, 1996/00

no abstracts in English

32 (Records 1-20 displayed on this page)