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Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi
Annals of Nuclear Energy, 138, p.107182_1 - 107182_9, 2020/04
Times Cited Count：0 Percentile：100(Nuclear Science & Technology)The investigation on self-shielding effect of double heterogeneity for plutonium burner High Temperature Gas-cooled Reactor (HTGR) design has been performed. Plutonium burner HTGR designed in the previous study by using the advantage of double heterogeneity to control excess reactivity. In the present study, the mechanism of the self-shielding effect is elucidated by the analysis of burn-up calculation and reactivity decomposition based on exact perturbation theory. As a result, it is revealed that the characteristics of burn-up reactivity are determined by resonance cross section peak at 1 eV of 
Pu due to the surface term of background cross section, this is, the characteristics of neutron leakage from fuel lump and collision to a moderator. Moreover, significant spectrum shift is caused during the burn-up period, and it enhances reactivity worth of 
Pu and Pu in EOL.
Sato, Hiroyuki; Aoki, Takeshi; Ohashi, Hirofumi; Yan, X.
Nuclear Engineering and Design, 360, p.110493_1 - 110493_8, 2020/04
Times Cited Count：0JAEA has been conducting research and development with a central focus on the utilization of HTTR, the first HTGR in Japan, towards the realization of industrial use of nuclear heat. On the basis of licensing experience through the HTTR construction, JAEA initiated an activity to establish an international safety standard for licensing of commercial HTGR cogeneration systems fully taking into account safety features of HTGRs. We have developed a roadmap towards licensing of commercial HTGR cogeneration systems. A test plan using the HTTR to support the establishment of safety standards and safety analysis methods are also presented. In addition, we confirmed that a vessel cooling system, a passive air-cooled decay heat removal system, satisfies the safety requirement.
Department of HTTR
JAEA-Review 2019-049, 97 Pages, 2020/03
JAEA-Review-2019-049.pdf:4.66MB
The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor being able to get 950
C temperature of the outlet coolant with 30MW of thermal power, constructed at the Oarai Research and Development Institute of the Japan Atomic Energy Agency is the first High- Temperature Gas-cooled Reactor (HTGR) in Japan. The purpose of the HTTR is to establish and upgrade basic technologies for HTGRs. The HTTR has accumulated a lot of experience of HTGRs' operation and maintenance up to the present time throughout rated power operations, safety demonstration tests, long-term high temperature operations and demonstration tests relevant to HTGRs' R&Ds. In the fiscal year 2018, we made effort to pass the inspection of application document for the HTTR licensing to prove conformity with the new regulatory requirements for research reactors that took effect since December 2013 in order to restart operations of the HTTR that stopped since the 2011 off the Pacific coast of Tohoku Earthquake. This report summarizes the activities carried out in the 2018 fiscal year, which were the situation of the new regulatory requirements screening of the HTTR, the operation and maintenance of the HTTR, R&Ds relevant to commercial-scale HTGRs, the international cooperation on HTGRs and so on.
Ueta, Shohei; Mizuta, Naoki; Fukaya, Yuji; Goto, Minoru; Tachibana, Yukio; Honda, Masaki*; Saiki, Yohei*; Takahashi, Masashi*; Ohira, Koichi*; Nakano, Masaaki*; et al.
Nuclear Engineering and Design, 357, p.110419_1 - 110419_10, 2020/02
Times Cited Count：0 Percentile：100(Nuclear Science & Technology)The concept of a plutonium (Pu) burner HTGR is proposed to incarnate highly-effective Pu utilization by its inherent safety features. The security and safety fuel (3S-TRISO fuel) employs the coated fuel particle with a fuel kernel made of plutonium dioxide (PuO
) and yttria stabilized zirconia (YSZ) as an inert matrix. This paper presents feasibility study of Pu burner HTGR and R&D on the 3S-TRISO fuel.
Aihara, Jun; Goto, Minoru; Ueta, Shohei; Tachibana, Yukio
JAEA-Data/Code 2019-018, 22 Pages, 2020/01
JAEA-Data-Code-2019-018.pdf:1.39MB
Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO (PuO-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. On the other hand, we have developed Code-B-2.5.2 for prediction of pressure vessel failure probabilities of SiC-tri-isotropic (TRISO) coated fuel particles for HTGRs under operation by modification of an existing code, Code-B-2. The main purpose of modification is preparation of applying code for CFPs of Pu-burner HTGR. In this report, basic formulae are described.
Aihara, Jun; Yasuda, Atsushi*; Ueta, Shohei; Ogawa, Hiroaki; Honda, Masaki*; Ohira, Koichi*; Tachibana, Yukio
Nippon Genshiryoku Gakkai Wabun Rombunshi, 18(4), p.237 - 245, 2019/12
Development of fabrication and inspection technologies of oxidation resistant fuel element for improvement of safety of high temperature gas-cooled reactors (HTGRs) in severe oxidation accident was carried out. Simulated coated fuel particles (CFPs), alumina particles, were over-coated with mixed powder of Si, C and small amount of resin to form over-coated particles, and over-coated particles were molded and hot-pressed to sinter simulated oxidation resistant fuel elements with SiC/C mixed matrix. Simulated oxidation resistant fuel elements with matrix whose Si/C mole ratio is 1.00 were fabricated. Failure fraction of CFPs in fuel elements is one of very important inspection subjects of HTGR fuel. It is essential that CFPs are extracted from fuel elements without additional failure. Development of method for extraction of CFPs was carried out. Desolation of SiC by KOH method or pressurized acidolysis method should be applied to extraction of CFPs.
Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*
Annals of Nuclear Energy, 133, p.830 - 836, 2019/11
Times Cited Count：0 Percentile：100(Nuclear Science & Technology)A RCCS having passive safety features through radiation and natural convection was proposed. The RCCS design consists of two continuous closed regions: an ex-reactor pressure vessel region and a cooling region with a heat-transfer surface to ambient air. The RCCS uses a novel shape to remove efficiently the heat released from the RPV through as much radiation as possible. Employing air as the working fluid and ambient air as the ultimate heat sink, the RCCS design can strongly reduce the possibility of losing the working fluid and the heat sink for decay-heat-removal. Moreover, the authors started experiment research with using a scaled-down heat-removal test facility. Therefore, this study propose a comparative methodology between an actual RCCS and a scaled-down heat-removal test facility.
Myagmarjav, O.; Tanaka, Nobuyuki; Nomura, Mikihiro*; Kubo, Shinji
International Journal of Hydrogen Energy, 44(59), p.30832 - 30839, 2019/11
Times Cited Count：0 Percentile：100(Chemistry, Physical)Ono, Masato; Kozawa, Takayuki; Fujimoto, Nozomu*
JAEA-Technology 2019-012, 15 Pages, 2019/09
JAEA-Technology-2019-012.pdf:2.83MB
The High Temperature Engineering Test Reactor has a neutron source of 
Cf to start up the reactor and to confirm count rates of wide range monitors. The half-life of Cf is short, about 2.6 years, so it is necessary to replace at an appropriate time. In order to estimate the period to replace, it is necessary to consider not only the half-life but also the fluctuation of the count rate of the wide range monitor to prevent alarm. For that reason, the method has been derived to predict a minimum count rate from relationship between the count rate and the standard deviation of the count rate of the wide range monitors. As a result of predicting the count rate using this method, it was found that the minimum count rate reaches to 3.0cps in 2022 and 1.5 cps in 2024. Therefore, it is necessary to exchange Cf by 2024.
Fukaya, Yuji; Nakagawa, Shigeaki; Goto, Minoru; Ishitsuka, Etsuo; Kawakami, Satoru; Uesaka, Takahiro; Morita, Keisuke; Sano, Tadafumi*
KURNS Progress Report 2018, P. 148, 2019/08
The Japan Atomic Energy Agency (JAEA) started the Research and Development (R&D) to improve nuclear prediction techniques for High Temperature Gas-cooled Reactors (HTGRs). The objectives are to introduce generalized bias factor method to avoid full mock-up experiment for the first commercial HTGR and to introduce reactor noise analysis to High Temperature Engineering Test Reactor (HTTR) experiment. To achieve the objectives, the reactor core of graphite moderation system named B7/4"G2/8"p8EUNU+3/8"p38EU(1) was newly composed in the B-rack of Kyoto University Critical Assembly (KUCA). The core plays a role of the reference core of the bias factor method, and the reactor noise was measured to develop the noise analysis scheme. In addition, training of operator of HTTR was also performed during the experiments.
Ogawa, Masuro*
JAEA-Technology 2019-010, 22 Pages, 2019/07
JAEA-Technology-2019-010.pdf:1.5MB
Transition phenomena from laminar to turbulent flow are roughly classified into three categories. Circular pipe flow of the third category is linearly stable against any small disturbance, despite that flow actually transitions and transitional flow exhibits intermittency. These are among major challenges that are yet to be resolved in fluid dynamics. Thus, author proposes hypothesis as follows; "Flow in a circular pipe transitions from laminar flow because of vortices released from separation bubble forming in vicinity of inlet of pipe, and transitional flow becomes intermittent because vortex-shedding is intermittent." Present hypothesis can easily explain why linear stability theory has not been able to predict transition in circular pipe flow, why circular pipe flow actually transitions, why transitional flow actually exhibits intermittency even due to small disturbance, and why numerical analysis has not been able to predict intermittency of transitional flow in circular pipe.
Ishitsuka, Etsuo; Matsunaka, Kazuaki*; Ishida, Hiroki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Kondo, Atsushi*; et al.
JAEA-Technology 2019-008, 12 Pages, 2019/07
JAEA-Technology-2019-008.pdf:2.37MB
As a summer holiday practical training 2018, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out. As a result, it is become clear that the continuous operations for about 30 years at 2 MW, about 25 years at 3 MW, about 18 years at 4 MW, about 15 years at 5 MW are possible. As an image of thermal design, the image of the nuclear battery consisting a cooling system with natural convection and a power generation system with no moving equipment is proposed. Further feasibility study to confirm the feasibility of nuclear battery will be carried out in training of next fiscal year.
Department of HTTR
JAEA-Review 2019-006, 97 Pages, 2019/07
JAEA-Review-2019-006.pdf:10.18MB
The High Temperature Engineering Test Reactor (HTTR) was constructed to establish and upgrade basic technologies for HTGRs. In the fiscal year 2017 we continued activities for re-operation of the HTTR and have been inspected the application document for the HTTR licensing to prove conformity with the new regulatory requirements for research reactors taken effect since December 2013 had been applied. This report summarizes activities and results of HTTR operation, maintenance, international cooperation and so on which were carried out in the fiscal year 2017.
Hirota, Noriaki; Kasahara, Seiji; Iwatsuki, Jin; Imai, Yoshiyuki; Ohashi, Hirofumi; Yan, X.; Tachibana, Yukio
Zairyo To Kankyo, 68(6), p.137 - 142, 2019/06
New corrosion test equipment for high temperature gas of decomposed sulfuric acid was manufactured in order to ascertain flow rate of sulfuric acid in the piping, occurrence of sulfuric acid decomposition reaction in the equipment, and temperature distribution inside the furnace tube. The flow rate of the sulfuric acid solution was constantly measured using an ultrasonic flowmeter. The SO
concentration at the inlet of the test equipment was almost the same as that at the inlet of the sulfuric acid decomposer in the hydrogen production plant assuming a high-temperature gas cooled reactor hydrogen-power cogeneration system (GTHTR300C). On the other hand, during a test, leakage of sulfuric acid occurred from the fitting part at the outlet side. Hence the temperature distribution of the fitting part at the outlet side was investigated using fluid analysis. As a result, it was found that the temperature at the fitting was low enough to use fluorine joint grease when the distance was 0.05 m or more away from the outlet side pipe. An improved furnace tube was manufactured and the temperature was measured again at fitting part of the outlet side. The temperature was lower that the temperature limit of the joint grease and almost the same as the temperature distribution in the fluid analysis, and leakage of sulfuric acid has not occurred to date.
Fukaya, Yuji
Meruko Kanri Kaikei Kenkyu, (11-2), p.45 - 62, 2019/05
Evaluation of electricity generation cost is important to make energy policy. Levelizing technique for cost and benefit, i.e. discounting technique, plays an important role in the evaluation especially for Nuclear Power Generation (NPG) because the activity related to power generation have been performed beyond generations. There are many arguments for the discounting technique for a long term, e.g. the cost of climate change. However, those are few for the electricity generation cost for NPG. Therefore, I discuss the discounting technique by referring the history of cost evaluation for NPG and the recent discussion of backend problem. Moreover, gamma discounting, which attracts a particular interest because of the gradual decrees of effective discounting rate, and two-good model are attempted and discussed.
Sato, Hiroyuki; Ohashi, Hirofumi
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
To establish a probabilistic approach for assessment of toxic gas leakage accidents in a H plant, the present study focusses on development of an uncertainty analysis method for toxic gas concentration in a control room. The method consists of 6 steps; (1) Identification of uncertainty factors, (2) derivation of variable parameters, (3) identification of uncertainties in variable parameters, (4) identification of important factors considering the sensitivity analysis results and expert opinions, (5) uncertainty propagation analysis, (6) assessment of uncertainty analysis results. The method is then applied to representative toxic gas leakage accidents in a H plant by IS process coupled to the HTTR. The results obtained in the study leads us to the conclusion that the suggested method can successfully characterize and quantify uncertainties in the toxic gas concentration in control room.
Takada, Shoji; Narayana, I. W.*; Nakatsuru, Yukihiro*; Terada, Atsuhiko; Murakami, Kenta*; Sawa, kazuhiko*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 13 Pages, 2019/05
In the loss of core cooling test using HTTR, a technical issue is to improve prediction accuracy of temperature distribution of components in vessel cooling system (VCS). An establishment of reasonable 2D model was started by using numerical code FLUENT, which was validated using the test data by 1/6 scale model of VCS for HTTR. The pressure vessel (PV) temperature was set around 200C attributed to relatively high ratio of natural convection heat transfer around 20% in total heat removal, which is useful for code to experiment benchmark to improve prediction accuracy. It is necessary to confirm heat transfer flow characteristics around the top of PV which is heated up by natural convection flow which was considered to be affected by separation, re-adhesion and transition flow. The k-
-SST model was selected for turbulent calculation attributed to predict the effects mentioned above adequately. The numerical results using the k--SST model reproduced the temperature distribution of PV especially the top region which is considered to be affected by separation, re-adhesion and transition flow in contract to that using k-
model which does not account the effects.
Ioka, Ikuo; Kuriki, Yoshiro*; Iwatsuki, Jin; Kubo, Shinji; Inagaki, Yoshiyuki; Sakaba, Nariaki
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 5 Pages, 2019/05
A thermochemical water-splitting iodine-sulfur processes (IS process) is one of candidates for the large-scale production of hydrogen using heat from solar power. Severe corrosive environment which is thermal decomposition of sulfuric acid exists in the IS process. A hybrid material with the corrosion-resistance and the ductility was made by a plasma spraying and laser treatment. The specimen had excellent corrosion resistance in the condition of 95 mass% boiling sulfuric acid. This was attributed to the formation of SiO on the surface. To confirm the production characteristics of a container using the hybrid material, the container which has a welded part, a chamfer, a curved surface was experimentally made. There was no detachment in the plasma spraying and laser treated layer of the container after the laser treatment.
Ueta, Shohei; Mizuta, Naoki; Sasaki, Koei; Sakaba, Nariaki; Ohashi, Hirofumi; Yan, X.
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
JAEA has been progressing to design HTGR fuels for not only small-type practical HTGRs but also VHTR proposed in GIF which can be utilized for various purposes with high-temperature heat at 750 to 950 C. To increase economy of these HTGRs, JAEA has been upgrading the design method for the HTGR fuel, which can maintain their integrities at the burnup of three to four times higher than that of the conventional HTTR fuel. Design principles and specifications of various concepts of the high burnup HTGR fuels designed by JAEA are reported. As the latest results on post-irradiation examinations of the high burnup HTGR fuel progressing in a framework of international collaboration with Kazakhstan, irradiation shrinkage rate of the fuel compact as a function of fast neutron fluence was obtained at around 100 GWd/thm. Furthermore, the future R&Ds needed for the high burnup HTGR fuel are described based on these experimental results.
Mizuta, Naoki; Aoki, Takeshi; Ueta, Shohei; Ohashi, Hirofumi; Yan, X.
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 5 Pages, 2019/05
Enhancement of safety and cooling performance of fuel elements are desired for a commercial High Temperature Gas-cooled Reactor (HTGR). Applying sleeveless fuel elements and dual side directly cooling structures with oxidation resistant SiC-matrix fuel compact has a possibility of improving safety and cooling performance at the pin-in-block type HTGR. The irradiated effective thermal conductivity of a fuel compact is an important physical property for core thermal design of the pin-in-block type HTGR. In order to discuss the irradiated effective thermal conductivity of the SiC-matrix fuel compact which could improve the cooling performance of the reactor, the maximum fuel temperature during normal operation of the pin-in-block type HTGR with dual side directly cooling structures are analytically evaluated. From these results, the desired irradiated thermal conductivity of SiC matrix are discussed. In addition, the suitable fabrication method of SiC-matrix fuel compact is examined from viewpoints of the sintering temperature, the purity and the mass productivity.
