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Hamamoto, Shimpei; Shimizu, Atsushi; Inoi, Hiroyuki; Tochio, Daisuke; Homma, Fumitaka; Sawahata, Hiroaki; Sekita, Kenji; Watanabe, Shuji; Furusawa, Takayuki; Iigaki, Kazuhiko; et al.
Nuclear Engineering and Design, 388, p.111642_1 - 111642_11, 2022/03
Times Cited Count：2 Percentile：53.91(Nuclear Science & Technology)Following the Fukushima Daiichi Nuclear Power Plant accident in 2011, the Japan Atomic Energy Agency adapted High-Temperature engineering Test Reactor (HTTR) to meet the new regulatory requirements that began in December 2013. The safety and seismic classifications of the existing structures, systems, and components were discussed to reflect insights regarding High Temperature Gas-cooled Reactors (HTGRs) that were acquired through various HTTR safety tests. Structures, systems, and components that are subject to protection have been defined, and countermeasures to manage internal and external hazards that affect safety functions have been strengthened. Additionally, measures are in place to control accidents that may cause large amounts of radioactive material to be released, as a beyond design based accident. The Nuclear Regulatory Commission rigorously and appropriately reviewed this approach for compliance with the new regulatory requirements. After nine amendments, the application to modify the HTTR's installation license that was submitted in November 2014 was approved in June 2020. This response shows that facilities can reasonably be designed to meet the enhanced regulatory requirements, if they reflect the characteristics of HTGRs. We believe that we have established a reference for future development of HTGR.
Takada, Shoji; Honda, Yuki*; Inaba, Yoshitomo; Sekita, Kenji; Nemoto, Takahiro; Tochio, Daisuke; Ishii, Toshiaki; Sato, Hiroyuki; Nakagawa, Shigeaki; Sawa, Kazuhiro*
Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10
Nuclear heat utilization systems connected to HTGRs will be designed on the basis of non-nuclear grade standards for easy entry of chemical plant companies, requiring reactor operations to continue even if abnormal events occur in the systems. The inventory control is considered as one of candidate methods to control reactor power for load following operation for siting close to demand area, in which the primary gas pressure is varied while keeping the reactor inlet and outlet coolant temperatures constant. Numerical investigation was carried out based on the results of nuclear heat supply fluctuation tests using HTTR by non-nuclear heating operation to focus on the temperature transient of the reactor core bottom structure by imposing stepwise fluctuation on the reactor inlet temperature under different primary gas pressures below 120C. As a result, it was emerged that the fluctuation absorption characteristics are not deteriorated by lowering pressure. It was also emerged that the reactor outlet temperature did not reach the scram level by increasing the reactor inlet temperature 10 C stepwise at 80% of the rated power as same with the full power case.
Tochio, Daisuke; Honda, Yuki; Sato, Hiroyuki; Sekita, Kenji; Homma, Fumitaka; Sawahata, Hiroaki; Takada, Shoji; Nakagawa, Shigeaki
Journal of Nuclear Science and Technology, 54(1), p.13 - 21, 2017/01
Times Cited Count：1 Percentile：10.62(Nuclear Science & Technology)GTHTR300C is designed and developed in JAEA. The reactor system is required to continue a stable and safety operation as well as a stable power supply in the case that thermal-load is fluctuated by the occurrence of abnormal event in the heat utilization system. Then, it is necessary to demonstrate that the thermal-load fluctuation should be absorbed by the reactor system so as to continue the stable and safety operation could be continued. The thermal-load fluctuation absorption tests without nuclear heating were planned and conducted in JAEA to clarify the absorption characteristic of thermal-load fluctuation mainly by the reactor and by the IHX. As the result it was revealed that the reactor has the larger absorption capacity of thermal-load fluctuation than expected one, and the IHX can be contributed to the absorption of the thermal-load fluctuation generated in the heat utilization system in the reactor system. It was confirmed from there result that the reactor and the IHX has effective absorption capacity of the thermal-load fluctuation generated in the heat utilization system. Moreover it was confirmed that the safety estimation code based on RELAP5/MOD3 can represents the thermal-load fluctuation absorption behavior conservatively.
Inaba, Yoshitomo; Sekita, Kenji; Nemoto, Takahiro; Honda, Yuki; Tochio, Daisuke; Sato, Hiroyuki; Nakagawa, Shigeaki; Takada, Shoji; Sawa, Kazuhiro
Journal of Nuclear Engineering and Radiation Science, 2(4), p.041001_1 - 041001_7, 2016/10
The nuclear heat utilization systems connected to High Temperature Gas-cooled Reactors (HTGRs) will be designed on the basis of non-nuclear grade standards in terms of the easier entry of chemical plant companies and the construction economics of the systems. Therefore, it is necessary that the reactor operations can be continued even if abnormal events occur in the systems. The Japan Atomic Energy Agency has developed a calculation code to evaluate the absorption of thermal load fluctuations by the reactors when the reactor operations are continued after such events, and has improved the code based on the High Temperature engineering Test Reactor (HTTR) operating data. However, there were insufficient data on the transient temperature behavior of the metallic core side components and the graphite core support structures corresponding to the fluctuation of the reactor inlet coolant temperature for further improvement of the code. Thus, nuclear heat supply fluctuation tests with the HTTR were carried out in non-nuclear heating operation to focus on thermal effect. In the tests, the coolant helium gas temperature was heated up to 120
C by the compression heat of the gas circulators in the HTTR, and a sufficiently high fluctuation of 17C by devising a new test procedure was imposed on the reactor inlet coolant under the ideal condition without the effect of the nuclear power. Then, the temperature responses of the metallic core side components and the graphite core support structures were investigated. The test results adequately showed as predicted that the temperature responses of the metallic components are faster than those of the graphite structures, and the mechanism of the thermal load fluctuation absorption by the metallic components was clarified.
Honda, Yuki; Tochio, Daisuke; Nakagawa, Shigeaki; Sekita, Kenji; Homma, Fumitaka; Sawahata, Hiroaki; Sato, Hiroyuki; Sakaba, Nariaki; Takada, Shoji
JAEA-Technology 2016-016, 16 Pages, 2016/08
JAEA-Technology-2016-016.pdf:2.84MB
A system analysis code is validated with the thermal-load fluctuation absorption test with nun-nuclear heating by using the High Temperature Engineering test Reactor (HTTR) to clarify the High Temperature Gas-cooled Reactor (HTGR) system response against temperature transient. The thermal-load fluctuation absorption test consists on the thermal load fluctuation tests (non-nuclear heating) and heat application system abnormal simulating test (non-nuclear heating). The HTGR reactor response against temperature transient is clarified in the thermal load fluctuation test (non-nuclear heating). The Intermediate Heat Exchanger (IHX) reactor response against temperature transient is clarified in the heat application system abnormal simulating test (non-nuclear heating). With the two HTTR non-nuclear heating test, HTGR system response against temperature transient is obtained.
Hamamoto, Shimpei; Nemoto, Takahiro; Sekita, Kenji; Saito, Kenji
JAEA-Technology 2015-048, 62 Pages, 2016/03
JAEA-Technology-2015-048.pdf:2.58MB
The decarburization may take place depending on the chemical impurity composition in helium gas used as the primary coolant in High-Temperature Gas-cooled Reactors, and will significantly reduce the strength of the alloy. The ability to remove impurities by a helium purification system was designed according to the predicted generation rate of impurities so as to make the coolant become the carburizing atmosphere. It has been confirmed that the coolant becomes the carburizing atmosphere during the operation period of the High Temperature engineering Test Reactor (HTTR). However, it is necessary to consider changes of generation rates of impurities since lifetime of commercial reactor is longer than the life of the HTTR. To avoid the influence of the change of generation rate, the control of removal efficiency of impurity in the helium purification system was considered in this study. To reform the decarburizing into the carburizing atmosphere, it is effective to increase the H
and CO concentration in the coolant helium. By controlling the efficiency of the Cooper Oxide Trap (CuOT), it is possible to increase the H and CO concentrations. Therefore, an experiment was carried out by injecting the gas mixture of H and CO into the existing purification system of HTTR to investigate the dependencies of temperature and impurity concentration on the removal efficiency of CuOT. The experimental results are described as the following, (1) By adjusting the temperature of helium at the CuOT within a range from 110C to 50C, it is possible to reduce the removal efficiency of H sufficiently. (2) Temperature change of helium gas in the CuOT is sufficiently reduced by the cooler located at the downstream of the CuOT, which does not affect the primary cooling system of HTTR. As the results, the applicability of removal efficiency control of CuOT was verified to improve the decarburizing atmosphere for the actual HTGR system.
Takada, Shoji; Sekita, Kenji; Nemoto, Takahiro; Honda, Yuki; Tochio, Daisuke; Inaba, Yoshitomo; Sato, Hiroyuki; Nakagawa, Shigeaki; Sawa, Kazuhiro
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05
To investigate the safety design criteria of heat utilization system for the HTGRs, it is necessary to evaluate the effect of fluctuation of thermal load on the reactor. The nuclear heat supply fluctuation test by non-nuclear heating was carried out to simulate the nuclear heat supply test which is carried out in the nuclear powered operation. The test data is used to verify the numerical code to calculate the temperature of core bottom structure to carry out the safety evaluation of abnormal events in the heat utilization system. In the test, the helium gas temperature was heated up to 120C. A sufficiently high temperature disturbance was imposed on the reactor inlet temperature. It was found that the response of temperatures of metallic components such as side shielding blocks was faster than those of graphite blocks in the core bottom structure, which was significantly affected by the heat capacities of components, the level of imposed disturbance and heat transfer performance.
Nemoto, Takahiro; Kaneshiro, Noriyuki*; Sekita, Kenji; Furusawa, Takayuki; Kuroha, Misao; Kawakami, Satoru; Kondo, Masaaki
JAEA-Technology 2015-006, 36 Pages, 2015/03
JAEA-Technology-2015-006.pdf:16.77MB
The High-Temperature engineering Test Reactor (HTTR) has been developed for establishing and upgrading the technical basis of HTGR.HTTR facilities have their structures, systems and a lot of components including reciprocating gas compressors, commonly used to extract and/or discharge reactor coolant helium gas contained in primary/secondary coolant systems. From the fact of the operational experiences of these compressors, seal-oil leakage has been frequently observed, although rod-seal mechanisms with complicated structures are equipped and improved for preventing coolant helium gas. So, we tried to change the rod-seal materials which might be a primary reason of frequent seal-oil leakage, that resulted in decreasing a mass and frequently of seal-oil leakage. It is confirmed that it is important to select adequate materials of rod seal for sliding speed of the piston of the compressor to prevent seal-oil leakage. Additionally, the procedure to estimate seal-oil leakage for each compressor is discussed. This report describes the results of investigation for improvement on seal-oil leak tightness of the compressors in HTTR facilities.
Tochio, Daisuke; Hamamoto, Shimpei; Inoi, Hiroyuki; Shimazaki, Yosuke; Sekita, Kenji; Kondo, Masaaki; Saikusa, Akio; Kameyama, Yasuhiko; Saito, Kenji; Emori, Koichi; et al.
JAEA-Technology 2010-038, 57 Pages, 2010/12
JAEA-Technology-2010-038.pdf:2.36MB
In HTTR, in-service operation is conducted through the rise-to power operation with rated operation or high-temperature test operation from achievement of first criticality at 1998. To make practical use HTGR system, it must be demonstrated to supply stable heat to heat utilization system for long-term. In HTTR, high-temperature/parallel-loaded long-term operation had been performed from January 2010. As the result, it was demonstrated to supply stable heat to heat utilization system for 50 days with HTTR, moreover, various long-term operation data were gained. This paper reports the characteristics of the high-temperature long-term operation for HTTR obtained from the operation.
Sekita, Kenji; Arakaki, Etsushi; Wakabayashi, Hiroshi
JAEA-Technology 2010-036, 42 Pages, 2010/11
JAEA-Technology-2010-036.pdf:2.9MB
The Liquid Nitrogen (LN) manufacturing facility in HTTR, is a high-pressure gas manufacturing facility. The facility provides LN to cold charcoal trap (CCT) in primary and secondary helium purification systems, impurities gas trap for gas chromatography mass spectrometer in primary and secondary helium sampling systems. The facility has been performed periodical inspection according to the law and rule, and each inspection item satisfied the judgment value. Thus, it was confirmed that the maintenance of LN manufacturing facility was appropriate. In addition, the effective and efficient LN amount to be consumed was clarified on the basis of the monitoring. Results of the liquid level of the storage tank and the consumption of LN as the reactor was operated or stopped, by considering the filling frequency and amount of the filling of LN to the facility by a tank lorry. This report describes method and result of periodical inspection and equipment management.
Tochio, Daisuke; Nojiri, Naoki; Hamamoto, Shimpei; Inoi, Hiroyuki; Sekita, Kenji; Kondo, Masaaki; Saikusa, Akio; Kameyama, Yasuhiko; Saito, Kenji; Fujimoto, Nozomu
JAEA-Technology 2009-005, 47 Pages, 2009/05
JAEA-Technology-2009-005.pdf:4.01MB
HTTR is now conducted in-service operation through the rise-to power operation with rated operation or high-temperature test operation from achievement of first criticality at 1998. In order to demonstrate to supply stable heat to heat utilization system for long-term, HTTR was conducted rated/parallel-loaded 30-days operation. This paper reports the characteristics of long-term operation for HTTR.
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.
Sekita, Kenji; Furusawa, Takayuki; Emori, Koichi; Ishii, Taro*; Kuroha, Misao; Hayakawa, Masato; Ouchi, Hiroshi
JAEA-Technology 2008-057, 45 Pages, 2008/08
JAEA-Technology-2008-057.pdf:12.0MB
A carbon steel used is used for the main material for the components and pipings of the pressurized water cooling system etc. that are the reactor cooling system of the HTTR. Water quality is managed by using the hydrazine in the coolant of the water cooling system to prevent corrosion of the components and deoxidize the coolant. Also, regular analysis is carried out for the confirmation of the water quality. The following results were obtained through the water quality analysis. (1) In the pressurized water cooling system, the coolant temperature rises higher due to the heat removal of the primary coolant. So, the ammonia was formed in the thermal decomposition of the hydrazine. The electric conductivity increased, while the concentration of the hydrazine decreased, there was no problem as the plan it. (2) Thermal decomposition of the hydrazine was not occurred in the auxiliary water cooling system and vessel cooling system because of the coolant temperature was low. (3) An indistinct procedure is clarified and procedure of water quality analysis was established in the HTTR. (4) It is assumed that the corrosion of the components in these water cooling system hardly occurred from measurement results of dissolved oxide and chloride ion. Thus, the water quality was managed enough.
Sekita, Kenji; Kuroha, Misao; Emori, Koichi; Kondo, Masaaki; Ouchi, Hiroshi; Shinozaki, Masayuki
JAEA-Technology 2008-002, 49 Pages, 2008/03
JAEA-Technology-2008-002.pdf:9.21MB
Graphite structures are used as one of the HTTR core internal structures. Graphite structures have high heat resistant property but its mechanical strength degrades easily by oxidization. To prevent the oxidization of graphite structures, impurity concentrations in the coolant of helium are controlled strictly. The helium sampling system is installed to measure the impurity concentrations in the helium. At gas compressor in helium sampling system, seal-oil leak at rod seal mechanism was occurred. The causes are degradation of seal material and contaminant abrasion powder of grand-packing. As these countermeasure, material of seal material was changed and contaminant was decreased. As the result long term operation is enabled. Moreover, reliable data can be obtained and efficient impurity control is enabled due to renewal of data acquisition control computer of gas chromatograph mass spectrometer and improvement of liquid nitrogen trap.
Saito, Kenji; Sekita, Kenji; Kawasaki, Kozo; Yamamoto, Kazuhiko*; Matsuura, Makoto*
JAEA-Technology 2007-059, 36 Pages, 2007/11
JAEA-Technology-2007-059.pdf:26.24MB
The Wide-Range Monitoring neutron detectors of HTTR are used under 450 C in normal operation and 550 C in the accidents. When the WRM detectors are used under the high temperature for a long time, characteristics of the detector might be degraded, because of the decrease of the nitrogen concentration in the ionization gas caused by adsorbtion of nitrogen into the electrode material. Consequently, the nitrogen gas adsorption test was carried out to clarify the quantity of absorbed nitrogen gas in electrode material under the high temperature. Then, the performance evaluation test of the prototype detector was carried out, and it was confirmed that degradation of the prototype detector characteristics didn't arise under the high temperature anvironment. This report describes the results of consideration about the life-extension of WRM detectors. As a result, it was confirmed that the WRM detectors are usable for 5 years under 450 C in normal operation and 550 C in the accidents.
Oyama, Sunao*; Hamamoto, Shimpei; Kaneshiro, Noriyuki*; Nemoto, Takahiro; Sekita, Kenji; Isozaki, Minoru; Emori, Koichi; Ito, Yoshiteru*; Yamamoto, Hideo*; Ota, Yukimaru; et al.
JAEA-Technology 2007-047, 40 Pages, 2007/08
JAEA-Technology-2007-047.pdf:18.83MB
High-Temperature engineering Test Reactor (HTTR) built by Japan Atomic Energy Agency (JAEA) has commonly used reciprocating compressor to extract helium gas and discharge helium gas into primary/secondary coolant helium loop from helium purification system. Rod-seal structure of the compressor is complicated from a prevention coolant leak standpoint. Because of frequently leakage of seal oil in operation, Rod seal structure isn't as reliable as it should be sustainable in the stable condition during long term operation. As a result of investigations, leakage's root is found in that seal were used in a range beyond limit sliding properties of seal material. Therefore a lip of the seal was worn and transformed itself and was not able to sustain a seal function. Endurance test using materials testing facility and verification test using a actual equipment on candidate materials suggest that a seal of fluorine contained resin mixed graphite is potentially feasible material of seal.
Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo
JAEA-Technology 2007-014, 62 Pages, 2007/03
JAEA-Technology-2007-014.pdf:9.74MB
The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was begun in April 2000. The reactor thermal power of 30 MW, which is the maximum thermal power of the HTTR, and the reactor outlet coolant temperature of 850C in normal operation was achieved in middle of December 2001. After that reactor thermal power of 30 MW a reactor outlet coolant temperature of 950C was achieved in the final rise-to-power test at April 2004. After receiving the operation permit, the safety demonstration tests were conducted to demonstrate inherent safety features of the HTGRs. This paper summarizes the HTTR operating experiences for five years since rise-to-power test that were catalogued into three categories, (1) Operating experience pertaining to new gas cooled reactor design, (2) Operating experience for improvement of the performance, (3) Operating experience due to fail of system and components.
Kondo, Masaaki; Sekita, Kenji; Emori, Koichi; Sakaba, Nariaki; Kimishima, Satoru; Kuroha, Misao; Noji, Kiyoshi; Aono, Tetsuya; Hayakawa, Masato
JAEA-Testing 2006-002, 55 Pages, 2006/07
JAEA-Testing-2006-002.pdf:6.36MB
The leakage rate test for the reactor containment vessel of HTTR is conducted in accordance with the absolute pressure method provided in Japan Electric Association Code(JEAC4203). Although leakage test of a reactor containment vessel is, in general, performed in condition of reactor coolant pressure boundary to be opened in order to simulate an accident, the peculiar test method to HTTR which use the helium gas as reactor coolant has been established, in which the pressure boundary is closed to avoid the release of fission products into the environment of the reactor containment vessel. The system for measuring and calculating the data for evaluating the leakage rate for containment vessel of HTTR was developed followed by any modifications. Recently, the system has been improved for more accurate and reliable one with any useful functions including real time monitoring any conditions related to the test. In addition, the configuration of containment vessel boundary for the test and the calibration method for the detectors for measuring temperature in containment vessel have been modified by reflecting the revision of the Code mentioned above. This report describes the method, system configuration, and procedures for the leakage rate test for reactor containment vessel of HTTR.
Aono, Tetsuya; Kondo, Masaaki; Sekita, Kenji; Emori, Koichi; Kuroha, Misao; Ouchi, Hiroshi
JAEA-Testing 2006-004, 39 Pages, 2006/06
JAEA-Testing-2006-004.pdf:9.88MB
The High Temperature Engineering Test Reactor (HTTR) has an emergency air purification system(EAPS). The system keeps the service area negative pressure condition and exhausts the filtered air to prevent fission products release to environment in accident condition. The EAPS is one of the engineered safety features which is started automatically when radioactivity in the service area increase or might increase. The performance of the EAPS should satisfy the analytical condition for public dose evaluation in the severest accidents of the HTTR. The performance should be confirmed by function tests. The function tests are divided into many tests corresponding to each assumed phenomenon. The confirmation of the performance of the system was carried out effectively by the tests. Moreover, the stable operation of the system can be achieved by improvements of the method of leak tight tests of exhaust filter unit. The report describes the outline of EAPS system, maintenance works and improvement of the system.
Sekita, Kenji; Emori, Koichi; Kuroha, Misao; Kimishima, Satoru; Wakabayashi, Hiroshi
JAEA-Testing 2006-001, 49 Pages, 2006/06
JAEA-Testing-2006-001.pdf:6.24MB
no abstracts in English
