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Takamatsu, Kuniyoshi
Annals of Nuclear Energy, 106, p.71 - 83, 2017/08
The HTTR, which is the only HTGR having inherent safety features in Japan, conducted a safety demonstration test involving a loss of both reactor reactivity control and core cooling. The paper shows thermal-hydraulics during the LOFC test at an initial power of 30% reactor power (9 MW), when the insertion of all control rods was disabled and all gas circulators were tripped to reduce the coolant flow rate to zero. The analytical results could show that the downstream of forced convection caused by the HPS pushes down the upstream by natural convection in the fuel assemblies; however, the forced convection has little influence on the core thermal-hydraulics without the reactor outlet coolant temperature. As a result, the three-dimensional thermal-phenomena inside the RPV during the LOFC test could be understood qualitatively.
Ono, Masato; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Takada, Shoji; Sawa, Kazuhiro
Journal of Nuclear Engineering and Radiation Science, 2(4), p.044502_1 - 044502_4, 2016/10
In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to verify safety evaluation codes to investigate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. The VCS passively removes the retained residual heat and the decay heat from the core via the reactor pressure vessel by natural convection and thermal radiation. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. Through a cold test, which was carried out by non-nuclear heat input from gas circulators with stopping water flow in the VCS, the local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1
C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.
Takada, Shoji; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Seki, Tomokazu; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Sawa, Kazuhiro
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to demonstrate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. The local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.
Takamatsu, Kuniyoshi; Tochio, Daisuke; Nakagawa, Shigeaki; Takada, Shoji; Yan, X.; Sawa, Kazuhiro; Sakaba, Nariaki; Kunitomi, Kazuhiko
Journal of Nuclear Science and Technology, 51(11-12), p.1427 - 1443, 2014/11
Times Cited Count：13 Percentile：69.72(Nuclear Science & Technology)In a safety demonstration test involving a loss of both reactor reactivity control and core cooling, HTGRs such as the HTTR, which is the only HTGR in Japan, demonstrate that the reactor power would stabilize spontaneously. In the test at an initial power of 30%, when the insertion of all control rods was disabled and all gas circulators were tripped to reduce the coolant flow rate to zero, a reactor transient was initiated and examined. The results confirmed that the reactor power would decrease immediately and become effectively zero.
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki
Nihon 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 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. 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.
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Iyoku, Tatsuo
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11) (CD-ROM), 12 Pages, 2005/10
Safety demonstration tests using the HTTR are in progress to verify the inherent safety features, to improve the safety design and the technologies for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping one or two out of three gas circulators is one of the safety demonstration tests. The reactor power safely becomes a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The SIRIUS code was developed to analyze reactor transient during the tests with reactor dynamics. This paper describes the validation of the SIRIUS code with the measured values of one and two gas circulators tripping test at 30% (9 MW). It was confirmed that the SIRIUS code was able to analyze the reactor transient within 10% during the tests. The result of this study and the way of resolving problems can be applied to development for not only the commercial HTGRs but also the Very High Temperature Reactor (VHTR) as one of the Generation IV reactors.
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki
JAERI-Data/Code 2005-003, 31 Pages, 2005/06
JAERI-Data-Code-2005-003.pdf:4.83MB
Safety demonstration tests using the High Temperature engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The TAC/BLOOST code was developed to analyze reactor and temperature transient during the coolant flow reduction test taking account of reactor dynamics. This paper describes the validation result of the TAC/BLOOST code with the measured values of gas circulators tripping tests at 30 % (9 MW). It was confirmed that the TAC/BLOOST code was able to analyze the reactor transient during the test.
Ono, Tomio*; Subekti, M.*; Kudo, Kazuhiko*; Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Nabeshima, Kunihiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 4(2), p.115 - 126, 2005/06
Control-rod withdrawal tests simulating reactivity insertion are carried out in the HTTR to verify the inherent safety features of HTGRs. This paper describes pre-test analysis method using artificial neural networks to predict the changes of reactor power and reactivity. The network model applied in this study is based on recurrent neural networks. The inputs of the network are the changes of the central control rods position and other significant core parameters, and the outputs are the changes of reactor power and reactivity. Furthermore, Time Synchronizing Signal(TSS) is added to input to improve the modeling of time series data. The actual tests data, which were previously carried out in the HTTR, were used for learning the model of the plant dynamics. After the learning, the network can predict the changes of reactor power and reactivity in the following tests.
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.
Tachibana, Yukio
Genshiryoku Nenkan 2005-Nen Ban, p.279 - 287, 2005/00
no abstracts in English
Takamatsu, Kuniyoshi; Furusawa, Takayuki; Hamamoto, Shimpei; Nakagawa, Shigeaki
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 11 Pages, 2004/10
Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping one or two out of three gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. Through the safety demonstration test, the two dimensional temperature analysis code (TAC-NC code) was improved. This paper describes the validation of the TAC-NC code using the measured value of the test by tripping of one and two out of three gas circulators at 30%(9MW). The TAC-NC code could evaluate accurately the temperature transient within 10% during the test. Also, it was confirmed that the temperature transient by tripping all gas circulators is very slow.
Tachibana, Yukio; Nakagawa, Shigeaki; Nakazawa, Toshio; Iyoku, Tatsuo
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 17 Pages, 2004/10
no abstracts in English
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 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
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.
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.
Tachibana, Yukio
Genshiryoku Nenkan 2004-Nen Ban, p.79 - 87, 2003/11
no abstracts in English
Tachibana, Yukio; Nakagawa, Shigeaki; Takeda, Takeshi; Saikusa, Akio; Furusawa, Takayuki; Takamatsu, Kuniyoshi; Sawa, Kazuhiro; Iyoku, Tatsuo
Nuclear Engineering and Design, 224(2), p.179 - 197, 2003/09
Times Cited Count：13 Percentile：64.52(Nuclear Science & Technology)no abstracts in English
Tachibana, Yukio; Nakagawa, Shigeaki; Iyoku, Tatsuo
Proceedings of International Conference on Global Environment and Advanced Nuclear Power Plants (GENES4/ANP 2003) (CD-ROM), 8 Pages, 2003/09
no abstracts in English
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.
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.
