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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：8.38(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.
Homma, Fumitaka; Hirato, Yoji; Saito, Kenji
JAEA-Technology 2014-010, 64 Pages, 2014/05
JAEA-Technology-2014-010.pdf:35.55MB
After an accident at Fukushima Daiichi Nuclear Plant, the High Temperature Engineering Test Reactor (HTTR) should be kept shut-down till adaptability to the new nuclear safety regulation established by the Nuclear Regulation Authority (NRA) will be confirmed. However, we have to keep the operational ability for the HTTR is indispensable. The operation training for the HTTR using the simulator of light water reactors is not effective. Because the HTTR is the only test reactor in the world and one has many characteristics different from light water reactors. It is an urgent problem to utilize the device which confirms the responsiveness of the control systems for the HTTR (simulator) as a simulator for operation training. This report shows specifications of device, results of simulation and the matter which we should carry out in future.
Yamanaka, Atsushi; Hashimoto, Kowa; Uchida, Toyomi; Shirato, Yoji; Isozaki, Toshihiko; Nakamura, Yoshinobu
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12
The Tokai Reprocessing Plant (TRP) adopted the PUREX method in 1977 and has reprocessed spent nuclear fuel of 1140 tHM (tons of heavy metals) since then. The reprocessing equipment suffers from various corrosion phenomena because of high nitric acidity, solution ion concentrations, such as uranium, plutonium, and fission products, and temperature. Therefore, considering corrosion performance in such a severe environment, stainless steels, titanium steel, and so forth were employed as corrosion resistant materials. The severity of the corrosive environment depends on the nitric acid concentration and the temperature of the solution, and uranium in the solution reportedly does not significantly affect the corrosion of stainless steels and controls the corrosion rates of titanium steel. The TRP equipment that handles uranyl nitrate solution operates at a low nitric acid concentration and has not experienced corrosion problems until now. However, there is a report that corrosion rates of some stainless steels increase in proportion to rising uranium concentrations. The equipment that handles the uranyl nitrate solution in the TRP includes the evaporators, which concentrate uranyl nitrate to a maximum concentration of about 1000 gU/L (grams of uranium per liter), and the denitrator, where uranyl nitrate is converted to UO
powder at about 320
C. These equipments are therefore required to grasp the degree of the progress of corrosion to handle high-temperature and high-concentration uranyl nitrate. The evaluation of this equipment on the basis of thickness measurement confirmed only minor corrosion and indicated that the equipment would be fully adequate for future operation.
Saito, Kenji; Shimizu, Atsushi; Hirato, Yoji; Kondo, Makoto; Kawamata, Takanori; Nemoto, Masumi; Motegi, Toshihiro
JAEA-Technology 2009-015, 52 Pages, 2009/05
JAEA-Technology-2009-015.pdf:10.17MB
As In-core temperature monitoring system, Type N thermocouples arranged at hot plenum block measures the primary coolant temperature at each hot plenum block in order to monitor the condition of the reactor core during the reactor operation. Type N thermocouples should have a long lifetime with high reliability under the high temperature environment of about 1000C, because they are used in HTTR reactor pressure vessel. This report shows that the characteristic change of Type N thermocouples was confirmed from operation and maintenance data of current HTTR.
Motegi, Toshihiro; Iigaki, Kazuhiko; Saito, Kenji; Sawahata, Hiroaki; Hirato, Yoji; Kondo, Makoto; Shibutani, Hideki; Ogawa, Satoru; Shinozaki, Masayuki; Mizushima, Toshihiko; et al.
JAEA-Technology 2006-029, 67 Pages, 2006/06
JAEA-Technology-2006-029.pdf:3.07MB
The plant control performance of the IHX helium flow rate control system, the PPWC helium flow rate control system, the secondary helium flow rate control system, the inlet temperature control system, the reactor power control system and the outlet temperature control system of the HTTR are obtained through function tests and power-up tests. As the test results, the control systems show stable control response under transient condition. Both of inlet temperature control system and reactor power control system shows stable operation from 30% to 100%, respectively. This report describes the outline of control systems and test results.
Saito, Kenji; Nakagawa, Shigeaki; Hirato, Yoji; Kondo, Makoto; Sawahata, Hiroaki; Tsuchiyama, Masaru*; Ando, Toshio*; Motegi, Toshihiro; Mizushima, Toshihiko; Nakazawa, Toshio
JAERI-Tech 2004-042, 26 Pages, 2004/04
JAERI-Tech-2004-042.pdf:1.16MB
The reactor control system of HTTR is composed of the reactor power control system, the reactor inlet coolant temperature control system, the primary coolant flow rate control system and so on. The reactor control system of HTTR achieves reactor power 30MW, reactor outlet coolant temperature 850C, reactor inlet coolant temperature 395C under the condition that primary coolant flow rate is fixed. In the Rise-to-Power Test, the performance test of the reactor inlet coolant temperature control system was carried out in order to confirm the control capability of this control system. This report shows the test results of performance test. As a result, the control parameters, which can control the reactor inlet coolant temperature stably during the reactor operation, were successfully selected. And it was confirmed that the reactor inlet coolant temperature control system has the capability of controlling the reactor inlet coolant temperature stably against any disturbances on the basis of operational condition of HTTR.
Hirato, Yoji; Saito, Kenji; Kondo, Makoto; Sawahata, Hiroaki; Motegi, Toshihiro; Tsuchiyama, Masaru*; Ando, Toshio*; Mizushima, Toshihiko; Nakazawa, Toshio
JAERI-Tech 2004-037, 33 Pages, 2004/04
JAERI-Tech-2004-037.pdf:4.08MB
HTTR (High Temperature Engineering Test Reactor) was operated from May 6th, 2003 to June 18th, 2003 to obtain operation data in parallel loaded operation mode and in safety demonstration tests. Operated with the reactor power at 60% of the rated power on May 21st, HTTR was automatically scrammed by a signalof "Primary coolant flow rate of the Primary Pressurized Water Cooler (PPWC): Low". The cause of the shutdown was the primary gas circulator (A) automatically stopped. The primary coolant flow rate of the PPWC decresed and reached the scram set value due to the gas circulator stop. As a result of investigation, it became clear that the cause of the gas circulator stop was malfunction of an auxiliary relay which monitored electric power of a circuit breaker in power line of the gas circulator. The cause of malfunction was deterioration of the relay under high temperature condition because the relay was installed beside an electric part which was heated up by electricity.
Shirato, Yoji; Isozaki, Toshihiko; Kishi, Yoshiyuki; Isobe, Hiroyasu; Nakamura, Yoshinobu; Uchida, Toyomi; Seno, Shigeo
no journal, ,
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Isozaki, Toshihiko; Tsutagi, Koichi; Shirato, Yoji; Nakazawa, Yutaka; Kake, Yasuhiro; Furukawa, Shinichi
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Kinuhata, Hiroshi*; Kodama, Takashi*; Nakano, Masanao*; Tamauchi, Yoshikazu*; Matsuoka, Shingo*; Tomiyama, Masahiro; Yasuda, Takeshi; Tsutagi, Koichi; Yoshino, Yasuyuki; Shirato, Yoji; et al.
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Kondo, Makoto; Homma, Fumitaka; Sawahata, Hiroaki; Hirato, Yoji; Kawamoto, Taiki; Suzuki, Hisashi; Ono, Masato
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Hirato, Yoji; Kozawa, Takayuki; Saito, Kenji
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Kishi, Yoshiyuki; Yasuda, Takeshi; Tokoro, Hayate; Yamanaka, Atsushi; Tsutagi, Koichi; Shirato, Yoji; Tanaka, Hitoshi
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High active liquid waste is stored in the facility equipped with cooling system and hydrogen scavenging system because of hydrogen generation due to radiolysis, and heat generation due to decay heat. The facilities related to these systems maintenance have been designed to be fed by emergency power generators in the conventional design. In addition to that, in light of the lessons learned from accident at the Fukushima Daiichi Nuclear Power Plant, we took safety measures for emergencies such as securement of the power supply system from power source vehicle to enable a quick recovery in the case of station black out in the Tokai Reprocessing Plant.
Kozawa, Takayuki; Hirato, Yoji; Homma, Fumitaka
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Murakami, Manabu; Yamanaka, Atsushi; Nakazawa, Yutaka; Goto, Yuichi; Shirato, Yoji; Uchida, Toyomi
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Isozaki, Toshihiko; Shirato, Yoji; Tsutagi, Koichi; Yoshino, Yasuyuki; Uchida, Toyomi; Nakamura, Yoshinobu
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Tomiyama, Masahiro; Yasuda, Takeshi; Tsutagi, Koichi; Yoshino, Yasuyuki; Shirato, Yoji; Nakamura, Yoshinobu; Kinuhata, Hiroshi*; Kodama, Takashi*; Nakano, Masanao*; Tamauchi, Yoshikazu*; et al.
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Honda, Yuki; Tochio, Daisuke; Aono, Tetsuya; Hirato, Yoji; Kozawa, Takayuki; Saito, Kenji
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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.
Murakami, Manabu; Omura, Masami; Furukawa, Shinichi; Shirato, Yoji; Ishiyama, Koichi
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Shirato, Yoji; Yamanaka, Atsushi; Tsutagi, Koichi; Yoshino, Yasuyuki; Kishi, Yoshiyuki; Isobe, Hiroyasu
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