Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.
High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02
As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.
Shimizu, Atsushi; Furusawa, Takayuki; Homma, Fumitaka; Inoi, Hiroyuki; Umeda, Masayuki; Kondo, Masaaki; Isozaki, Minoru; Fujimoto, Nozomu; Iyoku, Tatsuo
Journal of Nuclear Science and Technology, 51(11-12), p.1444 - 1451, 2014/11
JAEA has kept up a data-base system of operation and maintenance experiences of the HTTR. The objective of this system is to share the information obtained operation and maintenance experiences and to make use of lessons learned and knowledge into a design, construction and operation managements of the future HTGR. More than one thousand records have been registered into the system between 1997 and 2012. This paper describes the status of the data-base system, and provides suggestions for improvement from four experiences: (1) performance degradation of helium compressors; (2) malfunction of reserved shutdown system in reactivity control system; (3) maintenance experiences of emergency gas turbine generators; and (4) experiences of the Great East Japan Earthquake. These experiences are extracted from the system as important lessons learned to be expected to apply for design, construction and operation managements of future HTGR.
Takamatsu, Kuniyoshi; Sawa, Kazuhiro; Kunitomi, Kazuhiko; Hino, Ryutaro; Ogawa, Masuro; Komori, Yoshihiro; Nakazawa, Toshio*; Iyoku, Tatsuo; Fujimoto, Nozomu; Nishihara, Tetsuo; et al.
Nihon Genshiryoku Gakkai Wabun Rombunshi, 10(4), p.290 - 300, 2011/12
A high temperature (950C) continuous operation has been performed for 50 days on the HTTR from January to March in 2010, and the potential to supply stable heat of high temperature for hydrogen production for a long time was demonstrated for the first time in the world. This successful operation could establish technological basis of HTGRs and show potential of nuclear energy as heat source for innovative thermo-chemical-based hydrogen production, emitting greenhouse gases on a "low-carbon path" for the first time in the world.
Goto, Minoru; Fujimoto, Nozomu; Shimakawa, Satoshi; Tachibana, Yukio; Nishihara, Tetsuo; Iyoku, Tatsuo
Proceedings of 5th International Topical Meeting on High Temperature Reactor Technology (HTR 2010) (CD-ROM), 8 Pages, 2010/10
In the High Temperature Engineering Test Reactor (HTTR), which is a Japanese block-type HTGR, reactivity is controlled by control rods (CRs) and burnable poisons (BPs). The CRs insertion depth into the core should be retained shallow during burnup period, because the large insertion depth leads to significant disturbance of the power distribution, and consequently fuel temperature rises above the limit. Thus, the controllable reactivity with the CRs during operation is small, and then reactivity control through the burnup period largely depends on the BPs. It has not been confirmed an effectiveness of BPs on reactivity control on block-type HTGRs. The HTTR succeeded in long-term high temperature operation, and its burnup reached about 370EFPD. Thereby it became possible to confirm the effectiveness of BPs on reactivity control on the HTTR using its burnup data. We focused on a burnup change in the CRs insertion depth into the core to confirm whether the BPs functioned as designed. Additionally, we compared the change in the CRs insertion depths between analysis results and the experimental data to confirm validity of a whole core burnup calculation with the SRAC/COREBN. As a result, the experimental data showed that although the CRs insertion depth into the core was increased with burnup, it was retained the allowable depth. Meanwhile, the analysis result of the CRs insertion depth was in good agreement with the experimental data.
Shinohara, Masanori; Tochio, Daisuke; Hamamoto, Shimpei; Inoi, Hiroyuki; Shinozaki, Masayuki; Nishihara, Tetsuo; Iyoku, Tatsuo
Proceedings of 5th International Topical Meeting on High Temperature Reactor Technology (HTR 2010) (CD-ROM), 7 Pages, 2010/10
HTTR constructed at the Oarai Research and Development Center of JAEA is the first HTGR in Japan. The reactor thermal power is 30 MW, the reactor maximum outlet coolant temperature is 850 C in rated operation mode and 950 C in high temperature test operation mode. Main objectives of the HTTR are to establish and develop HTGR technology and to demonstrate process heat application. 30-days operation in rated operation mode and 50-days operation in high-temperature operation mode were performed to obtain various characteristic data of HTGR. The main test results are as follows :(1) CPF of the HTTR has excellent confinement ability of fission product which is the highest performance in the world. (2) The measurement temperature of the core internals is good agreement with the design value so that their structural integrity is maintained. (3) The intermediate heat exchanger keeps excellent heat transfer performance from beginning of operation.
Nakajima, Norihiro; Nishida, Akemi; Suzuki, Yoshio; Yamada, Tomonori; Takemiya, Hiroshi; Iyoku, Tatsuo
Proceedings of Joint International Conference of 7th Supercomputing in Nuclear Application and 3rd Monte Carlo (SNA + MC 2010) (USB Flash Drive), 4 Pages, 2010/10
FIESTA is a large scale simulation, which concerns a simulation space to bring real world in digital space, such as so named either virtual reality or virtual facility. In order to establish a huge and complex object like nuclear facilities of a real world, there are lack of methodology and technology for reproducing facilities in digital space. FIESTA attempts to realize an assembly structural analysis for supplying one of the methodologies to reproduce facilities in digital space. The first attempt of the structural analysis for assembly is accomplished by finite element analysis by integrating parts of facilities. Since the structural analysis for assembly requests massive calculation, parallel and distributed computing was applied for the computational environment. The structural analysis for assembly by finite element method is confirmed to be able to analyze a huge and complex facility and show results of numerical experiment by applying to a test reactor driven by JAEA.
Shibata, Taiju; Kunimoto, Eiji*; Eto, Motokuni*; Shiozawa, Shusaku; Sawa, Kazuhiro; Oku, Tatsuo*; Maruyama, Tadashi*
Journal of Nuclear Science and Technology, 47(7), p.591 - 598, 2010/07
no abstracts in English
Takada, Shoji; Nishihara, Tetsuo; Iyoku, Tatsuo; Nakazawa, Toshio; Komori, Yoshihiro
Nihon Genshiryoku Gakkai-Shi ATOMO, 52(7), P. 387, 2010/07
The 50-day long-term high-temperature operation was successfully attained by the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA), the rated thermal output of 30 MW and the maximum reactor outlet temperature of 950 C, first in the world. The operation was started on January 22 and accomplished on March 13 this year. Many data on the characteristics of reactor core physics and thermal hydraulics, the impurity control in coolant helium gas, the performance of high temperature components and the core internal structure integrity was acquired to establish the HTGR technology basis. The HTGR is expected as a green-house gas emission free heat source of innovative thermo-chemical hydrogen production system. It was demonstrated first in the world that high temperature gas can be stably supplied for long term period. In the next stage, the tests will be carried out to confirm the applicability and the extreme safety performance of HTGR by the HTTR.
Tachibana, Yukio; Nishihara, Tetsuo; Sakaba, Nariaki; Ohashi, Hirofumi; Sato, Hiroyuki; Ueta, Shohei; Aihara, Jun; Goto, Minoru; Sumita, Junya; Shibata, Taiju; et al.
JAEA-Technology 2009-063, 155 Pages, 2010/02
This report describes full scope of the feasible future test plan mainly using the HTTR. The test items cover fuel performance and radionuclide transport, core physics, reactor thermal hydraulics and plant dynamics, and reactor operations, maintenance, control, etc. The test results will be utilized for realization of Japan's commercial Very High Temperature Reactor (VHTR) system, GTHTR300C.
Shibata, Taiju; Eto, Motokuni*; Kunimoto, Eiji*; Shiozawa, Shusaku; Sawa, Kazuhiro; Oku, Tatsuo*; Maruyama, Tadashi*
JAEA-Research 2009-042, 119 Pages, 2010/01
For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a "Special committee on research on preparation for codes for graphite components in HTGR" at Atomic Energy Society of Japan (AESJ). As a result, "Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor" was established. In this draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. This draft standard is the first standard in the world which shows the concept of standard for the graphite core components in HTGR.
Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Iyoku, Tatsuo; Sawa, Kazuhiro
Journal of Nuclear Science and Technology, 46(7), p.690 - 698, 2009/07
Thermal conductivity of graphite components in HTGR is reduced by neutron irradiation. The reduced thermal conductivity is expected to be recovered by thermal annealing when irradiated graphite component is heated above irradiation temperature. In this study, the thermal conductivities of IG-110 graphite for the VHTR were measured systematically and thermal annealing effect was evaluated quantitatively. As the results, the thermal conductivities of IG-110 graphite were recovered up to 80% of unirradiated ones at maximum and the thermal annealing effect of IG-110 on thermal conductivity could be evaluated quantitatively using proposed thermal annealing evaluation model based on experimental results. Moreover, the calculated thermal conductivities of IG-110 with modified thermal resistance model were good agreement with experimental ones more than irradiation temperature. It implies that modified thermal resistance model can predict the thermal conductivity of IG-110.
Kunimoto, Eiji; Shibata, Taiju; Shimazaki, Yosuke; Eto, Motokuni*; Shiozawa, Shusaku; Sawa, Kazuhiro; Maruyama, Tadashi*; Oku, Tatsuo*
JAEA-Research 2009-008, 28 Pages, 2009/06
The VHTR is being focused and developed internationally. In Japan, the HTTR of the JAEA is in operation, and research and development for the development of commercial HTGRs are carried out. Nuclear graphites are used for core components of the HTGRs and expansion of irradiation data is necessary when enough irradiation data are not established, because the graphite components in the HTGRs are used at severer condition than that in the HTTR. The necessary database can be established by expansion of existing irradiation data with appropriate interpolation and extrapolation methods. This paper shows the reasonable interpolation and extrapolation method for IG-110 graphite which is used for the HTTR and a major candidate for the VHTR. The interpolation and extrapolation method was developed so as to be general by using the irradiation data of the other graphites. As a result, irradiation properties of the IG-110 graphite were successfully expanded to the VHTR condition for the first time and the irradiation properties being necessary for the design could be developed.
Tomimoto, Hiroshi; Umeda, Masayuki; Nishihara, Tetsuo; Iyoku, Tatsuo
UTNL-R-0471, p.11_1 - 11_9, 2009/03
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. HTTR reactor core consist of twelve kinds of enriched uranium fuel elements. Fuel rods were designed for avoiding fuel rod false loading because fuel rods number is 4770, and it was considered on handling. 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. Pre-service inspection was finished. And 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.
Shibata, Taiju; Sumita, Junya; Tada, Tatsuya; Hanawa, Satoshi; Sawa, Kazuhiro; Iyoku, Tatsuo
Journal of Nuclear Materials, 381(1-2), p.165 - 170, 2008/10
The lifetime extension of in-core graphite components is one of the key technologies for the VHTR. The residual stress in the graphite components caused by neutron irradiation at high temperatures affects their lifetime. Although oxidation damage in the components would not be significant in reactor normal operation, it should be checked as well. To evaluate the degradation of the graphite components directly by non-destructive way, the applicability of the micro-indentation and ultrasonic wave methods were investigated. The fine-grained isotropic graphites of IG-110 and IG-430, the candidate grades for the VHTR, were used in this study. The following results were obtained. (1) The micro-indentation behavior was changed by applying the compressive strain on the graphite. It suggested that the residual stress would be measured directly. (2) The change of ultrasonic wave velocity with 1 MHz by the uniform oxidation could be evaluated by the wave-propagation analysis with wave-pore interaction model. (3) The trend of oxidation-induced strength degradation on IG-110 was expressed by using the proposed uniform oxidation model. The importance of the un-uniformity consideration was indicated.
Oku, Tatsuo*; Kurumada, Akira*; Imamura, Yoshio*; Ishihara, Masahiro
Journal of Nuclear Materials, 381(1-2), p.92 - 97, 2008/10
Four grades of graphites and two grades of C/C composites were irradiated by argon ions with 175 MeV. The load-depth curves were obtained from the micro indentation tests before and after ion irradiation, and the apparent hardness and the hardness property parameters B and D (close rerating to the strength and Young's modulus, respectively) were investigated. As a result, it is found that the apparent hardness and the parameters B and D increase due to argon ion irradiation, and that these micro hardness properties of the carbon materials are able to be expressed as a function of dpa values, including neutron irradiation data.
Iyoku, Tatsuo; Nojiri, Naoki; Fujimoto, Nozomu; Shinohara, Masanori; Ota, Yukimaru; Tachibana, Yukio
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
The High Temperature Gas-cooled Reactor (HTGR) is expected to be one of the most promising energy sources not only for electricity generation and but also for process heat applications such as hydrogen production, desalination, etc. In Japan, since 1960s Japan Atomic Energy Agency (JAEA) has been developing HTGR technologies such fuel, high temperature metal, graphite, core physics, thermal hydraulics control and so forth. These technologies were well developed and used to design and construct the Japan's first HTGR, High Temperature Engineering Test Reactor (HTTR). It is a graphite-moderated and helium-cooled HTGR with the rated thermal power of 30 MW and the maximum outlet coolant temperature of 950C. The HTTR achieved the reactor outlet coolant temperature of 950C on April 19, 2004. It is the highest coolant temperature outside reactor pressure vessel in the world. So far, basic performance data of the HTTR during the power-up and long-term high temperature operation tests are accumulated. Except that, various unique tests concerning the HTGR safety are conducted to confirm inherent safety characteristics of the HTGR.
Tachibana, Yukio; Iyoku, Tatsuo; Sato, Hiroyuki; Kunitomi, Kazuhiko; Ogawa, Masuro
Proceedings of International Scientific-Practical Conference Nuclear Power Engineering in Kazakhstan, 10 Pages, 2008/06
JAEA (Japan Atomic energy Agency) has been designing a Small-sized Co-generation High Temperature Gas-cooled Reactor (HTGR), named High Temperature Reactor 50-Cogeneration (HTR50C), on the basis of design, construction and operational experience of the Japan's first HTGR, HTTR. The HTR50C that can be used not only for electric power generation but also for hydrogen production and district heating is considered one of the most promising nuclear reactors for countries where sufficient infrastructure such as power grids is not provided. Specification, equipment configuration, etc. of the HTR50C were determined, and also economic assessment was made.
Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Iyoku, Tatsuo; Sawa, Kazuhiro
JAEA-Research 2008-036, 33 Pages, 2008/03
To advance the performance and safety of HTGR, heat-resistant ceramic materials are expected to be used as reactor internals of HTGR. C/C composite and superplastic zirconia are the promising materials for this purpose. In order to use these new materials as reactor internals in HTGR, it is necessary to establish a structure design method to guarantee the structural integrity under environmental and load conditions. Therefore, C/C composite expected as reactor internals of VHTR is focused and an investigation on the structural design method applicable to the C/C composite and a basic applicability of the C/C composite to representative structures of HTGR were carried out in this report. As the results, it is found that the competing risk theory for the strength evaluation of the C/C composite is applicable to design method and C/C composite is expected to be used as reactor internals of HTGR.
Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Iyoku, Tatsuo; Sawa, Kazuhiro
JAEA-Research 2008-007, 30 Pages, 2008/03
Neutron irradiation remarkably reduces the thermal conductivity of graphite components in HTGR. The reduced thermal conductivity is expected to be recovered by annealing of irradiation-induced defects, when the graphite components are heated above the irradiation temperature. The annealing effect is not considered in the maximum fuel temperature analysis of the HTTR design from a viewpoint of conservative evaluation for the maximum fuel temperature. Therefore, it is expected that the temperature evaluation at accident conditions could be carried out more accurately with a reasonable stand point by considering the annealing effect. In order to advance the evaluation method for temperature analysis of accident in the HTGR, the annealing effect on thermal conductivity of graphite was evaluated quantitatively and the design curve on the thermal conductivity for graphite components of HTGR was proposed in this study.
Iyoku, Tatsuo; Nojiri, Naoki; Tochio, Daisuke; Mizushima, Toshihiko; Tachibana, Yukio; Fujimoto, Nozomu
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04
A HTGR is particularly attractive because of its capability of producing high temperature helium gas and its inherent safety characteristics. Hence, the HTTR wasconstructed at the Oarai Research Establishment of the Japan Atomic Energy Agency. The HTTR achieved the full power of 30MW and reactor outlet coolant temperature of about 850C on December 7, 2001. After several operation cycles, the HTTR achieved the reactor outlet coolant temperature of 950C on April 19, 2004. It is the highest coolant temperature outside reactor pressure vessel in the world. Extensive tests are planned in the HTTR and a process heat application system will be coupled to the HTTR, where hydrogen will be produced directly from the nuclear energy.