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.
Mineo, Hideaki; Nishihara, Tetsuo; Ohashi, Hirofumi; Goto, Minoru; Sato, Hiroyuki; Takegami, Hiroaki
Nihon Genshiryoku Gakkai-Shi ATOMO, 62(9), p.504 - 508, 2020/09
High-Temperature Gas-cooled Reactor (HTGR) is one of thermal neutron reactor-type that employs helium gas coolant and graphite moderator. It has excellent inherent safety and can supply high-temperature heat which can be used not only for electric power generation but also for a wide range of application such as hydrogen production. Therefore, HTGR is expected to be an effective technology for reducing greenhouse gases in Japan as well as overseas. In this paper, we will introduce the forefront of technological development that JAEA is working toward the realization of an HTGR system consisting of a high temperature gas reactor and heat utilization facilities such as gas-turbine power generation and hydrogen production.
Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi; Yan, X.; Nishihara, Tetsuo; Tsubata, Yasuhiro; Matsumura, Tatsuro
Journal of Nuclear Science and Technology, 55(11), p.1275 - 1290, 2018/11
To reduce environmental burden and thread of nuclear proliferation, multi-recycling fuel cycle with High Temperature Gas-cooled Reactor (HTGR) has been investigated. Those problems are solved by incinerating TRans Uranium (TRU) nuclides, which is composed of plutonium and Minor Actinoide (MA), and there is concept to realize TRU incineration by multi-recycling with Fast Breeder Reactor (FBR). In this study, multi-recycling is realized even with thermal reactor by feeding fissile uranium from outside of the fuel cycle instead of breeding fissile nuclide. In this fuel cycle, recovered uranium by reprocessing and natural uranium are enriched and mixed with recovered TRU by reprocessing and partitioning to fabricate fresh fuels. The fuel cycle was designed for a Gas Turbine High Temperature Reactor (GTHTR300), whose thermal power is 600 MW, including conceptual design of uranium enrichment facility. Reprocessing is assumed as existing Plutonium Uranium Redox EXtraction (PUREX) with four-group partitioning technology. As a result, it was found that the TRU nuclides excluding neptunium can be recycled by the proposed cycle. The duration of potential toxicity decaying to natural uranium level can be reduced to approximately 300 years, and the footprint of repository for High Level Waste (HLW) can be reduced by 99.7% compared with GTHTR300 using existing reprocessing and disposal technology. Suppress plutonium is not generated from this cycle. Moreover, incineration of TRU from Light Water Reactor (LWR) cycle can be performed in this cycle.
Nishihara, Tetsuo; Yan, X.; Tachibana, Yukio; Shibata, Taiju; Ohashi, Hirofumi; Kubo, Shinji; Inaba, Yoshitomo; Nakagawa, Shigeaki; Goto, Minoru; Ueta, Shohei; et al.
JAEA-Technology 2018-004, 182 Pages, 2018/07
Research and development on High Temperature Gas-cooled Reactor (HTGR) in Japan started since late 1960s. Japan Atomic Energy Agency (JAEA) in cooperation with Japanese industries has researched and developed system design, fuel, graphite, metallic material, reactor engineering, high temperature components, high temperature irradiation and post irradiation test of fuel and graphite, high temperature heat application and so on. Construction of the first Japanese HTGR, High Temperature engineering Test Reactor (HTTR), started in 1990. HTTR achieved first criticality in 1998. After that, various test operations have been carried out to establish the Japanese HTGR technologies and to verify the inherent safety features of HTGR. This report presents several system design of HTGR, the world-highest-level Japanese HTGR technologies, JAEA's knowledge obtained from construction, operation and management of HTTR and heat application technologies for HTGR.
Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi; Nishihara, Tetsuo; Tsubata, Yasuhiro; Matsumura, Tatsuro
Annals of Nuclear Energy, 116, p.224 - 234, 2018/06
Optimization of disposal method and scenario to reduce volume of High Level Waste (HLW) and the footprint in a geological repository for High Temperature Gas-cooled Reactor (HTGR) has been performed. It was found that HTGR has great advantages to reducing HLW volume and its footprint, which are high burn-up, high thermal efficiency and pin-in-block type fuel, compared with those of LWR and has potential to reduce those more in the previous study. In this study, the scenario is optimized, and the geological repository layout is designed with the horizontal emplacement based on the KBS-3H concept instead of the vertical emplacement based on KBS-3V concept employed in the previous study. As a result, for direct disposal, the repository footprint can be reduced by 20 % by employing the horizontal without change of the scenario. By extending 40 years for cooling time before disposal, the footprint can be reduced by 50 %. For disposal with reprocessing, the number of canister generation can be reduced by 20 % by extending cooling time of 1.5 years between the discharge and reprocessing. The footprint per electricity generation can be reduced by 80 % by extending 40 years before disposal. Moreover, by employing four-group partitioning technology without transmutation, the footprint can be reduced by 90 % with cooling time of 150 years.
Fukaya, Yuji; Goto, Minoru; Nishihara, Tetsuo
Nuclear Engineering and Design, 326, p.108 - 113, 2018/01
Burn-up characteristics and criticality of impurity contained into graphite structure for commercial scale prismatic High Temperature Gas-cooled Reactor (HTGR) have been investigated. For HTGR, of which the core is filled graphite structure, the impurity contained into the graphite has unignorable poison effect for criticality. Then, GTHTR300, commercial scale HTGR, employed high grade graphite material named IG-110 to take into account the criticality effect for the reflector blocks next to fuel blocks. The fuel blocks, which should also employ IG-110, employ lower grade graphite material named IG-11 from the economic perspective. In this study, the necessity of high grade graphite material for commercial scale HTGR is reconsidered by evaluating the burn-up characteristics and criticality of the impurity. The poison effect of the impurity, which is used to be expressed by a boron equivalent, reduces exponentially like burn-up of B, and saturate at a level of 1 % of the initial value of boron equivalent. On the other hand, the criticality effect of the boron equivalent of 0.03 ppm, which corresponds to a level of 1 % of IG-11 shows ignorable values lower than 0.01 %k/kk' for both of fuel blocks and reflector blocks. The impurity can be represented by natural boron without problem. Therefore, the poison effect of the impurity is evaluated with whole core burn-up calculations. As a result, it is concluded that the impurity is not problematic from the viewpoint of criticality for commercial scale HTGR because it is burned clearly until End of Cycle (EOC) even with the low grade graphite material of IG-11. According to this result, more economic electricity generation with HTGR is expected by abolishing the utilization of IG-110.
Fukaya, Yuji; Goto, Minoru; Nishihara, Tetsuo
JAEA-Research 2015-023, 44 Pages, 2016/02
A study on innovative High Temperature Gas-cooled Reactor (HTGR) to reduce generation of potential radiotoxicity had been performed. Unlike the Fast Breeder Reactor (FBR) and Accelerated Driven System (ADS), which can confine radioactive nuclides into its fuel cycle as multi-recycling and transmute, in this study we attempt to reduce the generation of the radiotoxicity itself by preventing the generation of Pu and MA, which is generated with the energy generation. In this context, we proposed the innovative HTGR that employs the Highly Enriched Uranium (HEU) fuel by removing U : source of the Pu and MA. However, there are the problems of fuel integrity, nuclear proliferation, nuclear self-regulation characteristics, and economy of electricity generation which are caused by employing HEU. For these problems, we investigated and proposed the solutions. Especially for the nuclear self-regulation characteristics, which were improved by adding Er, the optimized nuclear design was quantitatively determined and elucidated by the Bondarenko approach. As a result, it was confirmed that the proposed reactor can solves these problem for employing HEU fuel and the high specification and economy as same as those of standard HTGR fueled uranium.
Fukaya, Yuji; Goto, Minoru; Nishihara, Tetsuo
Nuclear Engineering and Design, 293, p.30 - 37, 2015/11
The investigation on the erbium loading method to improve reactivity coefficients for Low Radiotoxic Spent Fuel High Temperature Gas-cooled Reactor (LRSF-HTGR) is performed. The fuel employs HEU to reduce toxicity generation from uranium-238. The reactivity coefficients show positive values without any additive. Then, the erbium is loaded in the core to obtain negative reactivity coefficient due to the large resonance peak of neutron capture reaction of erbium-167. The loading methods are investigated. The erbium is mixed into fuel kernel of CPF, loaded by binary packing with fuel particle and erbium particle, and embedded into the graphite shaft deployed center of fuel compact. It is found that the erbium loading causes negative reactivity as a moderator temperature reactivity, and it should be loaded into fuel pin elements for pin-in-block type fuel from the viewpoint of heat transfer. Moreover, the erbium should be incinerated slowly to obtain negative reactivity coefficient even at EOC. The loading method which effectively causes self-shielding should be selected to avoid to be incinerated with burn-up. The mechanism is elucidated by application of Bondarenko approach. As a result, it is conclude that the erbium embedded into graphite shaft is preferable for LRSF-HTGR to remain the reactivity coefficient negative at EOC.
Fukaya, Yuji; Goto, Minoru; Nishihara, Tetsuo
JAEA-Technology 2015-017, 61 Pages, 2015/07
A development on nuclear design model for detailed design of Clean Burn HTGR had been performed. In the previous study, the fuel composition assumed in the study of Deep Burn proposed by GA in U.S. was employed. In the present study, that is estimated to reflect the situation of the nuclear fuel cycle in Japan. And, the evaluation method is refined from the viewpoint of traceability and a guarantee of quality. The deployment of control rod columns is investigated. Moreover, the Er loading is also investigated to obtain negative temperature coefficients in all range of operation. This model is developed for MVP code, which solves neutron transportation by Monte Carlo method. Validation of burn-up chain is also performed to adopt very high burn-up calculation. The core design is performed by COREBN code, which solves neutron diffusion equation by deterministic approach. Thus, the convertor which converts the cross section of MVP to COREBN is also developed.
Haneklaus, N.*; Reyes, R.*; Lim, W. G.*; Tabora, E. U.*; Palattao, B. L.*; Petrache, C.*; Vargas, E. P.*; Kunitomi, Kazuhiko; Ohashi, Hirofumi; Sakaba, Nariaki; et al.
Philippine Journal of Science, 144(1), p.69 - 79, 2015/06
The Philippines may profit from extracting uranium (U) from phosphoric acid during fertilizer production in a way that the recovered U can be beneficiated and taken as raw material for nuclear reactor fuel. Used in a high temperature reactor (HTR) that provides electricity and/or process heat for fertilizer processing and U extraction, energy-neutral fertilizer production, an idea first proposed by Haneklaus et al., is possible. This paper presents a first case study of the concept regarding a representative phosphate fertilizer plant in the Philippines and exemplary HTR designs (HTR50S and GTHTR300C) developed by the Japan Atomic Energy Agency (JAEA). Three different arrangements (version I-III), ranging from basic electricity supply to overall power supply including on site hydrogen production for ammonia conversion, are introduced and discussed.
Maeda, Koji; Sasaki, Shinji; Kumai, Misaki; Sato, Isamu; Suto, Mitsuo; Osaka, Masahiko; Goto, Tetsuo*; Sakai, Hitoshi*; Chigira, Takayuki*; Murata, Hirotoshi*
Journal of Nuclear Science and Technology, 51(7-8), p.1006 - 1023, 2014/07
Since the start of the severe accident at the Fukushima Daiichi Nuclear Power Plant in March 2011, concrete surfaces within the reactor buildings have been exposed to radioactive contaminants. Released radiation sources still remain too high to permit entry into some areas of the RBs to allow the damage to be assessed and to allow carrying out the restoration of lost safety functions, decommissioning activities, etc. In order to clarify the situation of this contamination in the RBs, 18 samples were subjected to analyses to determine the surface radionuclide concentrations and to characterize the radionuclide distributions in the samples. Decontamination tests on the sample of Unit 2 were conducted to reduce the levels of radioactivity present near the sample surface. As a result of the tests, the level of radioactivity of the sample was reduced with the removal of 97% of the contamination present near the sample surface.
Takada, Shoji; Iigaki, Kazuhiko; Shinohara, Masanori; Tochio, Daisuke; Shimazaki, Yosuke; Ono, Masato; Yanagi, Shunki; Nishihara, Tetsuo; Fukaya, Yuji; Goto, Minoru; et al.
Nuclear Engineering and Design, 271, p.472 - 478, 2014/05
JAEA has carried out research and development to establish the technical basis of HTGRs using HTTR. To connect hydrogen production system to HTTR, it is necessary to ensure the reactor dynamics when thermal-load of the system is lost. Thermal-load fluctuation test is planned to demonstrate the reactor dynamics stability and to validate plant dynamics codes. It will be confirmed that the reactor become stable state during losing a part of removed heat at heat-sink. A temperature coefficient of reactivity is one of the important parameters for core dynamics calculations, and changes with burnup because of variance of fuel compositions. Measurement of temperature coefficient of reactivity has been conducted to confirm the validity of calculated temperature coefficient of reactivity. A LOFC test using HTTR has been carried out to verify the inherent safety under the condition of LOFC while the reactor shut-down system disabled.
Maeda, Koji; Sasaki, Shinji; Kumai, Misaki; Sato, Isamu; Osaka, Masahiko; Fukushima, Mineo; Kawatsuma, Shinji; Goto, Tetsuo*; Sakai, Hitoshi*; Chigira, Takayuki*; et al.
Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.272 - 277, 2013/09
Takada, Shoji; Iigaki, Kazuhiko; Shinohara, Masanori; Tochio, Daisuke; Shimazaki, Yosuke; Ono, Masato; Nishihara, Tetsuo; Fukaya, Yuji; Goto, Minoru; Tachibana, Yukio; et al.
Proceedings of 6th International Topical Meeting on High Temperature Reactor Technology (HTR 2012) (USB Flash Drive), 8 Pages, 2012/10
JAEA has carried out research and development to establish the technical basis of HTGRs using HTTR. LOFC test to verify the inherent safety of HTGR under the condition of loss of forced cooling while the reactor shut-down system disabled was initiated. A temperature coefficient of reactivity is one of the important parameters for core dynamics calculations for safety analysis, and changes with burnup because of variance of fuel compositions, which has been measured to confirm the validity of the calculated ones. In order to connect hydrogen production system to HTTR, it is necessary to ensure the reactor safety when thermal-load of the hydrogen production system is lost. Thermal load fluctuation test is planned to demonstrate the reactor safety and gain the test data for validation of the plant dynamics code. It will be confirmed that the reactor become stable state during a part of removed heat at HTTR heat-sink is lost.
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.
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.
Goto, Minoru; Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Ueta, Shohei; Hamamoto, Shimpei; Ohashi, Hirofumi; Furusawa, Takayuki; Saito, Kenji; Shimazaki, Yosuke; Nishihara, Tetsuo
JAEA-Technology 2009-053, 48 Pages, 2009/10
Preliminary studies on the HTTR (High Temperature engineering Test Reactor) tests were conducted to obtain characteristics and demonstration data which were required to develop commercial HTGRs (high temperature gas-cooled reactors). The tests proposed in this study are as follows: nuclear heat supply characteristics tests, burned core tests, reactivity insertion tests, safety demonstration tests, fuel characteristics tests, annular core tests, fuel failure tests, tritium measurement tests, and health confirmation tests of high temperature equipments. Requirements for a development of commercial HTGRs and confirmation methods of the requirements by the HTTR tests were summarized. Preliminary analyses were performed for the burned core test and the safety demonstration test to obtain prediction data, which is compared with experimental data. Additionally, a feasibility analysis was performed on four types annular cores, which is composed of the HTTR's fresh fuels, from the point of view of shutdown margin and excess reactivity.
Nakagawa, Shigeaki; Tochio, Daisuke; Shinohara, Masanori; Nojiri, Naoki; Nishihara, Tetsuo; Goto, Minoru; Takamatsu, Kuniyoshi
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9476_1 - 9476_6, 2009/05
Sakaba, Nariaki; Tachibana, Yukio; Shimakawa, Satoshi; Ohashi, Hirofumi; Sato, Hiroyuki; Yan, X.; Murakami, Tomoyuki; Ohashi, Kazutaka; Nakagawa, Shigeaki; Goto, Minoru; et al.
JAEA-Technology 2008-019, 57 Pages, 2008/03
The small-sized and safe cogeneration High Temperature Gas-cooled Reactor (HTGR) 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 developing countries where sufficient infrastructure such as power grids is not provided. Thus, the small-sized cogeneration HTGR, named High Temperature Reactor 50-Cogeneration (HTR50C), was studied assuming that it should be constructed in developing countries. Specification, equipment configuration, etc. of the HTR50C were determined, and economical evaluation was made. As a result, it was shown that the HTR50C is economically competitive with small-sized light water reactors.
Goto, Yoshitaka*; Tanabe, Tetsuo*; Ishimoto, Yuki*; Masaki, Kei; Arai, Takashi; Kubo, Hirotaka; Tsuzuki, Kazuhiro*; Miya, Naoyuki
Journal of Nuclear Materials, 357(1-3), p.138 - 146, 2006/10
no abstracts in English