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Kozawa, Takayuki; Suganuma, Takuro; Homma, Fumitaka; Higashimura, Keisuke*; Ukai, Takayuki*; Saito, Kenji
JAEA-Technology 2023-007, 24 Pages, 2023/06
JAEA-Technology-2023-007.pdf:2.24MB
To improve the reliability of the HTTR wide range monitor in a high-temperature environment, structural changes of the wide range monitor were investigated. It was clear that the structure for directly joins of the MI cable core wire and metal tube instead of the joins with lead wire is the most reliable method with shortest way. From a result of the thermal cycle tests and high temperature endurance tests for a mock-up connecting this connection parts, it was clear that the soundness of the connection part was maintained under the usage conditions of the HTTR.
Ono, Masato; Kozawa, Takayuki; Fujimoto, Nozomu*
JAEA-Technology 2019-012, 15 Pages, 2019/09
JAEA-Technology-2019-012.pdf:2.83MB
The High Temperature Engineering Test Reactor has a neutron source of 
Cf to start up the reactor and to confirm count rates of wide range monitors. The half-life of Cf is short, about 2.6 years, so it is necessary to replace at an appropriate time. In order to estimate the period to replace, it is necessary to consider not only the half-life but also the fluctuation of the count rate of the wide range monitor to prevent alarm. For that reason, the method has been derived to predict a minimum count rate from relationship between the count rate and the standard deviation of the count rate of the wide range monitors. As a result of predicting the count rate using this method, it was found that the minimum count rate reaches to 3.0cps in 2022 and 1.5 cps in 2024. Therefore, it is necessary to exchange Cf by 2024.
Honda, Yuki; Saito, Kenji; Tochio, Daisuke; Aono, Tetsuya; Hirato, Yoji; Kozawa, Takayuki; Nakagawa, Shigeaki
Journal of Nuclear Science and Technology, 51(11-12), p.1387 - 1397, 2014/11
Times Cited Count：1 Percentile：8.88(Nuclear Science & Technology)The operational experiments of the HTTR would be useful for future high-temperature gas-cooled reactors (HTGRs). Main PID control constants of the HTTR are selected with reasonably damped characteristics and without undershoot or overshoot. For utilization the HTGR as a commercial reactor, it should be demonstrated that the HTGR system can supply stable heat to a heat utilization system for the long-term operation. The control characteristics in the long-term high-temperature operation are evaluated by the result of operation performed in 2010. In addition, from a viewpoint of HTGRs with heat utilization system, a future possibility of the experiments for heat utilization design is examined.
Kobayashi, Hidekazu; Amamoto, Ippei; Yokozawa, Takuma; Yamashita, Teruo; Nagai, Takayuki; Kitamura, Naoto*; Takebe, Hiromichi*; Mitamura, Naoki*; Tsuzuki, Tatsuya*
Proceedings of 15th International Conference on Environmental Remediation and Radioactive Waste Management (ICEM 2013) (CD-ROM), 6 Pages, 2013/09
no abstracts in English
as the dominant sludge constituent generated from the treatment of water used for cooling the stricken power reactorsAmamoto, Ippei; Kobayashi, Hidekazu; Yokozawa, Takuma; Yamashita, Teruo; Nagai, Takayuki; Kitamura, Naoto*; Takebe, Hiromichi*; Mitamura, Naoki*; Tsuzuki, Tatsuya*
Proceedings of 15th International Conference on Environmental Remediation and Radioactive Waste Management (ICEM 2013) (CD-ROM), 8 Pages, 2013/09
The great amount of water used for cooling the stricken power reactors at Fukushima Dai-ichi has resulted in accumulation of "remaining water". As the remaining water is subsequently contaminated by FPs, etc., it is necessary to decontaminate it in order to reduce the volume of liquid radioactive waste and to reuse it again for cooling the reactors. Various techniques are being applied to remove FP, etc. and to make stable waste forms. One of the methods using the iron phosphate glass as a medium is being developed to stabilize the strontium-bearing sludge whose main component is BaSO. From the results hitherto, the iron phosphate glass is regarded as a potential medium for the target sludge.
Tomimoto, Hiroshi; Kato, Yasushi; Owada, Hiroyuki; Sato, Nao; Shimazaki, Yosuke; Kozawa, Takayuki; Shinohara, Masanori; Hamamoto, Shimpei; Tochio, Daisuke; Nojiri, Naoki; et al.
JAEA-Technology 2009-025, 29 Pages, 2009/06
JAEA-Technology-2009-025.pdf:21.78MB
The first driver fuel of the HTTR (High Temperature Engineering test Reactor) was loaded in 1998 and the HTTR reached first criticality state in the same year. The HTTR has been operated using the first driver fuel for a decade. In Fuel elements assembling, 4770 of fuel rods which consist of 12 kinds of enrichment uranium are loaded into 150 fuel graphite blocks for HTTR second driver fuel elements. Measures of prevention of fuel rod miss loading, are employed in fuel design. Additionally, precaution of fuel handling on assembling are considered. Reception of fuel rods, assembling of fuel elements and storage of second driver fuels in the fresh fuel storage rack in the HTTR were started since June, 2008. Assembling, storage and pre-service inspection were divided into three parts. The second driver fuel assembling was completed in September, 2008. This report describes concerns of fuel handling on assembling and storage work for the HTTR fuel elements.
Ueta, Shohei; Tobita, Tsutomu*; Sawa, Kazuhiro; Tomimoto, Hiroshi; Kozawa, Takayuki; Inoi, Hiroyuki; Umeda, Masayuki
JAEA-Research 2008-096, 34 Pages, 2009/01
JAEA-Research-2008-096.pdf:10.12MB
The temperature monitors for fuel blocks in high temperature gas reactors during operation are being developed. The temperature monitors consist of alloy wires, with various melting points, sealed in quarts capsules. The temperature can be evaluated in the range from 600 to 1400 
C with 22 types of the temperature monitors. The temperature monitors have been irradiated by the capsule in JMTR, and then, PIEs such as X-ray radiograph and EPMA have been carried out. As the results of the PIE, it was estimated that the temperature monitors can be used up to 90 days at 1100 C, or up to 50 days at 1300
1350 C.
Kondo, Masaaki; Kimishima, Satoru*; Emori, Koichi; Sekita, Kenji; Furusawa, Takayuki; Hayakawa, Masato; Kozawa, Takayuki; Aono, Tetsuya; Kuroha, Misao; Ouchi, Hiroshi
JAEA-Technology 2008-062, 46 Pages, 2008/10
JAEA-Technology-2008-062.pdf:11.62MB
The reactor containment of HTTR is tested to confirm leak-tight integrity of itself. "Type A test" has been conducted in accordance with the standard testing method in JEAC4203 since the preoperational verification of the containment was made. Type A tests are identified as basic one for measuring containment leakage rate, it costs much, however. Therefore, the test program for HTTR was revised to adopt an efficient and economical alternatives including "Type B and Type C tests". In JEAC4203-2004, following requirements are specified for adopting alternatives: upward trend of leakage rate by Type A test due to aging should not be recognized; criterion of combined leakage rate with Type B and Type C tests should be established; the criteria for Type A test and combined leakage rate test should be satisfied; correlation between the leakage rates by Type A test and combined leakage rate test should be recognized. Considering the performances of the tests, the policies of corresponding to the requirements were developed, which were accepted by the regulatory agency. This report presents an outline of the tests, identifies issues on the conventional test and summarizes the policies of corresponding to the requirements and of implementing the tests based on the revised program.
Yokozawa, Takuma; Kobayashi, Hidekazu; Yamashita, Teruo; Nagai, Takayuki; Tokizawa, Takayuki; Naganuma, Masaki; Aoki, Katsumi
no journal, ,
no abstracts in English
Kobayashi, Hidekazu; Yokozawa, Takuma; Yamashita, Teruo; Nagai, Takayuki; Amamoto, Ippei; Ayame, Yasuo
no journal, ,
no abstracts in English
Amamoto, Ippei; Kobayashi, Hidekazu; Yokozawa, Takuma; Yamashita, Teruo; Nagai, Takayuki; Suzuki, Yoshikazu*; Takebe, Hiromichi*; Mitamura, Naoki*; Tsuzuki, Tatsuya*
no journal, ,
The great amount of water used for cooling the stricken reactors at Fukushima Dai-ichi Nuclear Power Plant following the earthquake and tsunami of 11 March 2011 had resulted in accumulation of "remaining water" in some buildings. From public announcements, it seems likely the decontamination process of La Hague reprocessing plant would be employed as one of the treatment processes for the remaining water contaminated by FPs such as Cs, Sr, etc.. Based on literature study, Cs would precipitate with ferrocyanide compound as well as Sr with BaSO by applying the above-mentioned process. In this study, BaSO as the simulated sludge was loaded into the iron phosphate glass (IPG) medium under different melting temperature. Based on the results, the performance of IPG containing BaO (decomposed from BaSO) improved with increase loading of BaO up to 48 mol% at lower melting temperature.
Amamoto, Ippei; Yokozawa, Takuma; Yamashita, Teruo; Nagai, Takayuki; Nakamura, Hirotaka*; Takebe, Hiromichi*; Mitamura, Naoki*; Tsuzuki, Tatsuya*; Kobayashi, Hidekazu
no journal, ,
no abstracts in English
Nagai, Takayuki; Oyama, Koichi; Kobayashi, Hidekazu; Morikawa, Yo; Yokozawa, Takuma; Nakajima, Masayoshi; Yamashita, Teruo; Amamoto, Ippei
no journal, ,
no abstracts in English
Amamoto, Ippei; Kobayashi, Hidekazu; Yokozawa, Takuma; Yamashita, Teruo; Nagai, Takayuki; Suzuki, Yoshikazu*; Takebe, Hiromichi*; Mitamura, Naoki*; Tsuzuki, Tatsuya*
no journal, ,
no abstracts in English
Yokozawa, Takuma; Kobayashi, Hidekazu; Yamashita, Teruo; Nagai, Takayuki; Amamoto, Ippei
no journal, ,
no abstracts in English
Kobayashi, Hidekazu; Yokozawa, Takuma; Yamashita, Teruo; Nagai, Takayuki; Amamoto, Ippei
no journal, ,
no abstracts in English
Kobayashi, Hidekazu; Amamoto, Ippei; Yokozawa, Takuma; Yamashita, Teruo; Nagai, Takayuki; Kitamura, Naoto*; Takebe, Hiromichi*; Mitamura, Naoki*; Tsuzuki, Tatsuya*
no journal, ,
no abstracts in English
by the IPG mediumAmamoto, Ippei; Kobayashi, Hidekazu; Yokozawa, Takuma; Yamashita, Teruo; Nagai, Takayuki; Kitamura, Naoto*; Takebe, Hiromichi*; Mitamura, Naoki*; Tsuzuki, Tatsuya*
no journal, ,
no abstracts in English
Kobayashi, Hidekazu; Amamoto, Ippei; Yokozawa, Takuma; Yamashita, Teruo; Nagai, Takayuki; Kitamura, Naoto*; Takebe, Hiromichi*; Mitamura, Naoki*; Tsuzuki, Tatsuya*
no journal, ,
no abstracts in English
Honda, Yuki; Tochio, Daisuke; Aono, Tetsuya; Hirato, Yoji; Kozawa, Takayuki; Saito, Kenji
no journal, ,
The HTGR system with helium turbine and hydrogen production system have been designed based on the HTTR experiments. For the future HTGR control system design, the thermal-load fluctuation and lost test for heat utilization system design is planned. The tests would be important for demonstration of the reactor stability despite thermal load fluctuation, detailed design of the HTGRs with heat utilization system and validation of plant dynamic codes for future HTGR system. Preliminary studies with all main control system are carried out to study on the control characteristics with the thermal-load fluctuation test and determine test condition on normal operation state such as the amount of decrease in the flow rate of the ACL.
