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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.
Inoi, Hiroyuki; Shimizu, Atsushi; Kameyama, Yasuhiko; Kobayashi, Shoichi; Shinozaki, Masayuki; Ota, Yukimaru; Kubo, Tsukasa*; Emori, Koichi
JAEA-Technology 2009-048, 48 Pages, 2009/10
JAEA-Technology-2009-048.pdf:5.0MB
The emergency power feeders of the High Temperature Engineering Test Reactor (HTTR) have gas turbine generators which are composed of gas turbin engines, generators and current breakers. The gas turbine generators have been overhauled and maintained to keep the performance. The maintenance technology was upgraded by improving their parts and surveillance method on the basis of the operational and maintenance experience. It can be clarified that the deterioration levels and the sudden deterioration timing are judged at an early stage by measuring the max exhaust temperature at the time of start in addition to check the starting time of the Gas Turbine Engines.
Kondo, Makoto; Iigaki, Kazuhiko; Motegi, Toshihiro; Emori, Koichi
JAEA-Testing 2009-002, 50 Pages, 2009/08
JAEA-Testing-2009-002.pdf:2.44MB
High Temperature engineering Test Reactor (HTTR) is a test reactor that build on sand layer formed during quaternary era. In principle, nuclear power generation facilities need to be supported on rigid rock formed during tertiary era, and now it is investigated to construct nuclear power generation facilities on sand layer formed during quaternary era for expanding suitable location. For this reason, we installed seismic observation systems in HTTR. And we have performed seismic safety evacuation using measured seismic acceleration in ground and building. This report compiles specifications of seismic observation systems, contents of check for seismic observation systems, check result of seismic observation systems and maintenance of seismic observation systems in 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.
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.
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.
Noji, Kiyoshi; Kameyama, Yasuhiko; Emori, Koichi; Aono, Tetsuya
JAEA-Testing 2006-003, 47 Pages, 2006/06
JAEA-Testing-2006-003.pdf:7.43MB
The FFD (Fuel Failure Detection) System has been installed in the HTTR in order to detects the abnormal release of fission products from the fuel during the operations. The FFD system samples the primary coolant from the high-temperature plenum division of the reactor core divided into seven regions. The system detects short life fission product(FP) gases from each region. The damaged region can be specified by the FFD system. In the design, it was considered that the change in the sampling flow rate during operation was not necessary. However, it became clear that the measured value became unstable because of a fluctuation of the sampling flow rate due to change in the primary coolant pressure during operation. Moreover, it was difficult to change the sampling flow rate during operation. The sampling flow rate was controlled by manual valves located in the service area where the entry is limited during operation. Therefore, an improvement was carried out to control the sampling flow rate from the outside of the service area. The stable measured value was obtained by the improvement. Moreover, noise reduction, improvement of oil level gauge of compressors gives excellent operation of the FFD. This report summarizes the maintenance work of detectors (precipitator), equipment and improvement items 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
Sakaba, Nariaki; Nakagawa, Shigeaki; Furusawa, Takayuki*; Emori, Koichi; Tachibana, Yukio
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(4), p.388 - 395, 2004/12
Chemistry control is important for the helium coolant of High Temperature Gas-cooled Reactors (HTGRs) because impurities cause oxidation of the graphite used in the core and corrosion of high temperature materials used in the heat exchanger. In the High Temperature Engineering Test Reactor (HTTR) which is the first HTGR in Japan, the chemical impurity concentration is restricted and its behaviour is monitored during reactor operations. The impurity is reduced by the helium purification system and the concentration is measured by the helium sampling system installed to the primary and secondary helium system, continuously. This paper describes the impurity behaviour during the rise-to-power test which is the initial power-up of the HTTR. Also, the amount of the emitted impurity to the primary circuit from the graphite component and insulator used at the concentric hot gas duct are evaluated. During the power up, any abnormal impurity increases were not obtained and the chemical composition of the primary circuit is sufficiently in the stability area to avoid carbon deposition.
Sakaba, Nariaki; Iigaki, Kazuhiko; Kondo, Masaaki; Emori, Koichi
Nuclear Engineering and Design, 233(1-3), p.135 - 145, 2004/10
Times Cited Count：5 Percentile：35.25(Nuclear Science & Technology)The containment structures of the HTTR consist of the reactor containment vessel, the service area, and the emergency air purification system, which minimise the release of fission products in postulated accidents which lead to fission product release from the reactor facilities. The reactor containment vessel is designed to withstand the temperature and pressure transients and to be leak-tight in the case of a rupture of the primary concentric hot gas duct, etc. The pressure inside the service area is maintained at a negative pressure by the emergency air purification system. The emergency air purification system will also remove airborne radio-activity and will maintain a correct pressure in the service area. The leak-tightness characteristics of the containment structures are described in this paper. The measured leakage rates of the reactor containment vessel were enough less than the specified leakage limit of 0.1%/d confirmed during the commissioning tests and annual inspections. The service area was kept the design pressure well below its allowable limitation by the emergency air purification system which filter efficiency of particle removal and iodine removal were well over the limited values. The obtained data demonstrates that the reactor containment structures were fabricated to minimise the release of fission products in the postulated accidents with fission product release from the reactor facilities.
Ueta, Shohei; Emori, Koichi; Tobita, Tsutomu*; Takahashi, Masashi*; Kuroha, Misao; Ishii, Taro*; Sawa, Kazuhiro
JAERI-Research 2003-025, 59 Pages, 2003/11
JAERI-Research-2003-025.pdf:2.53MB
In the safety design requirements for the High Temperature Engineering Test Reactor (HTTR) fuel, it is determined that "the as-fabricated failure fraction shall be less than 0.2%" and "the additional failure fraction shall be small through the full service period". Therefore the failure fraction should be quantitatively evaluated during the HTTR operation. In order to measure the primary coolant activity, primary coolant radioactivity signals the in safety protection system, the fuel failure detection (FFD) system and the primary coolant sampling system are provided in the HTTR. The fuel and fission product behavior was evaluated based on measured data in the rise-to-power tests (1) to (4). The measured fractional releases are constant at 2
10
up to 60% of the reactor power, and then increase to 710 at full power operation. The prediction shows good agreement with the measured value. These results showed that the release mechanism varied from recoil to diffusion of the generated fission gas from the contaminated uranium in the fuel compact matrix.
Ueta, Shohei; Sumita, Junya; Emori, Koichi; Takahashi, Masashi*; Sawa, Kazuhiro
Journal of Nuclear Science and Technology, 40(9), p.679 - 686, 2003/09
Times Cited Count：13 Percentile：64.66(Nuclear Science & Technology)no abstracts in English
Takamatsu, Kuniyoshi; Nakazawa, Toshio; Furusawa, Takayuki; Homma, Fumitaka; Saito, Kenji; Kokusen, Shigeru; Kamata, Takashi; Ota, Yukimaru; Ishii, Yoshiki; Emori, Koichi
JAERI-Tech 2003-062, 94 Pages, 2003/06
JAERI-Tech-2003-062.pdf:26.47MB
no abstracts in English
Tachibana, Yukio; Hontani, Koji*; Kojima, Takao; Takeda, Takeshi; Emori, Koichi; Saruta, Toru; Iyoku, Tatsuo; Kunitomi, Kazuhiko
JAERI-Tech 2000-026, p.61 - 0, 2000/03
JAERI-Tech-2000-026.pdf:2.18MB
no abstracts in English
Sakaba, Nariaki; Emori, Koichi; Saruta, Toru
JAERI-Tech 99-072, p.125 - 0, 1999/10
no abstracts in English
Takeda, Tetsuaki; Takenaka, Satsuki*; Hishida, Makoto;
Nihon Genshiryoku Gakkai-Shi, 37(10), p.948 - 958, 1995/00
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
Takenaka, Satsuki*; Takeda, Tetsuaki; Hishida, Makoto; *;
JAERI-M 94-024, 60 Pages, 1994/03
no abstracts in English
