Fukaya, Yuji; Ueta, Shohei; Goto, Minoru; Ohashi, Hirofumi
Annals of Nuclear Energy, 151( ), p.107937_1 - 107937_9, 2021/02
Feasibility study on Burnable Poison (BP) credit concept to High Temperature Gas-cooled Reactor (HTGR) fuel fabrication has been performed. By mixing BP into fuel material in the first place of fuel fabrication, criticality safety is ensured in the all fuel fabrication process even with high enrichment fuel such as 14 wt% used in commercial HTGR. However, the poison effect also prevents the criticality even in the HTGR core, and it may shorten cycle length and achievable burn-up of the core. Therefore, the effect is evaluated by whole core burn-up calculation. As a BP, boron, gadolinium, erbium, and hafnium are investigated. As a result, it is found that boron and gadolinium suit this concept and the 14 wt% fuel can be fabricated in the plant fabricating 9.9 wt% High Temperature engineering Test Reactor (HTTR) fuel. With the boron and gadolinium, the commercial HTGR fuel can be fabricated with the safety measure as same as Light Water Reactor (LWR) fuel facility to treat the fuel with the enrichment up to 5 wt%. Especially, gadolinium is significantly suitable to this concept due to the dependency to spectrum, and more enhanced safety measure is feasible as well.
Sato, Hiroyuki; Aoki, Takeshi; Ohashi, Hirofumi
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 10 Pages, 2020/08
The present study aims to propose a guidance that facilitates to determine fuel design limits of commercial HTGR on the basis of licensing experience through the HTTR construction. The guidance consists of a set of FOMs and a process to determine their evaluation criteria. The FOMs are firstly identified to satisfy safety requirements and a basic concept of safety guides established in a special committee under the AESJ with the support of the Research Association of High Temperature Gas Cooled Reactor Plant. The development process for the evaluation criteria takes into account not only the top-level regulatory criteria but also design dependent constraints including the performance of fission product containment in physical barriers other than fuel, fuel qualification criteria, design specifications of an instrumentation and control system. As a result, a comprehensive and transparent procedure for designers of prismatic-type commercial HTGR has been developed.
Aoki, Takeshi; Sato, Hiroyuki; Ohashi, Hirofumi
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08
The flow distribution analysis, which is a part of thermal hydraulic design of the prismatic-type of the high temperature gas cooled reactor (HTGR) considering unintended flows between graphite blocks, has been performed for steady and conservative conditions. On the other hand, the transient analysis for satisfactorily realistic conditions will be helpful for the design improvement of prismatic-type HTGR. The present study aims to develop the transient flow distribution analysis code and confirm its applicability for the transient flow distribution analysis for prismatic-type HTGRs during anticipated operational occurrences and accidents utilizing experiences on high temperature engineering test reactor (HTTR) design. The calculation model and code were developed and validated for analysis of the unintended flows in the core and the molecular diffusion dominant in beginning air ingress behavior in an air ingress accident.
Aoki, Takeshi; Isaka, Kazuyoshi; Sato, Hiroyuki; Ohashi, Hirofumi
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08
The flow distribution analysis performed in the HTGR design has to take into account the interaction thermal and radiation deformations of the graphite structure, and the gaps between the graphite structures forming unintended flow. In the present study, a user-friendly flow network calculation code (FNCC) has been developed on the basis of experiences of High Temperature engineering Test Reactor (HTTR) design for HTGR design with enhanced compatibility with other HTGR design codes and with considering graphite block deformation in iteration process without manual control. The validation of FNCC was performed for the one-column flow distribution test. The analytical results using FNCC showed good agreement with the experimental results. It is concluded that FNCC was validate for the analysis of distributions of flowrate and pressure for the flow network model including the unintended flow paths in prismatic-type HTGRs.
Sato, Hiroyuki; Ohashi, Hirofumi
Mechanical Engineering Journal (Internet), 7(3), p.19-00332_1 - 19-00332_11, 2020/06
An uncertainty analysis method for control room habitability under toxic gas leakage accidents in cogeneration HTGR is proposed to support risk-informed design of the plant. The method is applied to representative toxic gas leakage accidents in a IS process hydrogen production plant coupled to the HTTR gas turbine test plant. Epistemic and aleatory uncertainties for each variable parameter are identified and are propagated using Latin hypercube sampling. The analyses show that the suggested method can successfully characterize and quantify uncertainties in the toxic gas concentration in control room. The results lead us to the conclusion that toxic gas dispersion behavior analysis should combine two evaluation methods: dense gas dispersion model and computational fluid dynamics simulation.
Saito, Hiroyuki*; Machida, Akihiko*; Iizuka, Riko*; Hattori, Takanori; Sano, Asami; Funakoshi, Kenichi*; Sato, Toyoto*; Orimo, Shinichi*; Aoki, Katsutoshi*
Scientific Reports (Internet), 10, p.9934_1 - 9934_8, 2020/06
Neutron powder diffraction profiles were collected for iron deuteride (FeDx) while the temperature decreased from 1023 to 300 K for a pressure range of 4-6 GPa. The ' deuteride with a double hexagonal close-packed (dhcp) structure, which coexisted with other stable or metastable deutrides at each temperature and pressure condition, formed solid solutions with a composition of FeD at 673 K and 6.1 GPa and FeD at 603 K and 4.8 GPa. Upon stepwise cooling to 300 K, the D-content x increased to a stoichiometric value of 1.0 to form monodeuteride FeD. In the dhcp FeD at 300 K and 4.2 GPa, dissolved D atoms fully occupied the octahedral interstitial sites, slightly displaced from the octahedral centers in the dhcp metal lattice, and the dhcp sequence of close-packed Fe planes contained hcp-stacking faults at 12%. Magnetic moments with 2.11 0.06 B/Fe-atom aligned ferromagnetically in parallel on the Fe planes.
Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi
Annals of Nuclear Energy, 138, p.107182_1 - 107182_9, 2020/04
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. L.
Nuclear Engineering and Design, 360, p.110493_1 - 110493_8, 2020/04
JAEA 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
The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor being able to get 950C 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.
Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki
Nippon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00353_1 - 19-00353_6, 2020/03
Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium.
Fukaya, Yuji; Goto, Minoru; Nakagawa, Shigeaki; Nakajima, Kunihiro*; Takahashi, Kazuki*; Sakon, Atsushi*; Sano, Tadafumi*; Hashimoto, Kengo*
Proceedings of International Conference on the Physics of Reactors; Transition To A Scalable Nuclear Future (PHYSOR 2020) (USB Flash Drive), 8 Pages, 2020/03
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 a 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 observe subcriticality. 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 is composed of the fuel assembly, driver fuel assembly, graphite reflector, and polyethylene reflector. The fuel assembly is composed of enriched uranium plate, natural uranium plate and graphite plates to realize the average fuel enrichment of HTTR and it's spectrum. However, driver fuel assembly is necessary to achieve the criticality with the small-sized core. 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 this study, the overview of the criticality experiments is reported. The reactor configuration with graphite moderation system is rare case in the KUCA experiments, and this experiment is expected to contribute not only for an HTGR development but also for other types of a reactor in the graphite moderation system such as a molten salt reactor development.
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.
Aihara, Jun; Ueta, Shohei; Honda, Masaki*; Mizuta, Naoki; Goto, Minoru; Tachibana, Yukio; Okamoto, Koji*
Journal of Nuclear Materials, 522, p.32 - 40, 2019/08
In order to realize Pu-burner high temperature gas-cooled reactor (HTGR), coated fuel particles (CFPs) with PuO-yittria stabilized zirconia (YSZ) fuel kernel coated with ZrC is employed for high nuclear proliferation resistance and very high burn-up. Japan Atomic Energy Agency (JAEA) have carried out ZrC coatings of particles which simulated PuO-YSZ kernels (CeO-YSZ particles or commercially available YSZ particles). Ce was used as simulating element of Pu. In this manuscript, microstructures of ZrC coated CeO-YSZ or YSZ particles were reported.
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.
Inagaki, Yoshiyuki; Sakaba, Nariaki; Tanaka, Nobuyuki; Nomura, Mikihiro*; Sawada, Shinichi*; Yamaki, Tetsuya*
Nippon Kaisui Gakkai-Shi, 73(4), p.194 - 202, 2019/08
The thermochemical IS process is a promising hydrogen production method which can produce hydrogen in a large amount and stably with high efficiency by thermal splitting of water. Research and development on chemical reaction technology with membranes was conducted for the purpose of improving the efficiency of IS process and application of solar heat. The basic technology of ceramic membranes applied to decomposition reactions of hydrogen iodine and sulfuric acid was developed, and it is expected that the conversion rate on decomposition in each reaction can be remarkably improved. The basic technology of a cation exchange membrane applied to Bunsen reaction was developed with radiation-induced grafting technique, it is expected that the amount of iodine can be reduced to about one-fifth compared to the conventional method. These achievements are important technologies for practical use of the IS process.
Machida, Akihiko*; Saito, Hiroyuki*; Hattori, Takanori; Sano, Asami; Funakoshi, Kenichi*; Sato, Toyoto*; Orimo, Shinichi*; Aoki, Katsutoshi*
Scientific Reports (Internet), 9(1), p.12290_1 - 12290_9, 2019/08
Hexagonal close-packed iron hydride, hcp FeHx, is absent from the conventional phase diagram of the Fe-H system, although hcp metallic Fe exists stably over extensive temperature () and pressure () conditions, including those corresponding to the Earth's inner core. X-ray and neutron diffraction measurements at temperatures ranging from 298 to 1073 K and H pressures ranging from 4 to 7 GPa revealed that the hcp hydride was formed for FeH compositions when . Hydrogen atoms occupied the octahedral interstitial sites of the host metal lattice both partially and randomly. The hcp hydride exhibited a H-induced volume expansion of 2.48(5) /H-atom, which was larger than that of the face-centered cubic (fcc) hydride. The hcp hydride showed an increase in with , whereas the fcc hydride showed a corresponding decrease. The present study provides guidance for further investigations of the Fe-H system over an extensive -- region.
Department of HTTR
JAEA-Review 2019-006, 97 Pages, 2019/07
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.
Yoneda, Yasuhiro; Taniguchi, Hiroki*; Kitanaka, Yuki*; Noguchi, Yuji*
Ferroelectrics, 538(1), p.57 - 62, 2019/05
High-energy X-ray diffraction study was performed on lead titanate (PbTiO). Short-range order structure was revealed using atomic pair-distribution function (PDF) method. In the high-temperature cubic phase, there is a large deviation between local and average structure. Especially, Pb atoms deviated from the cubic lattice framework, but establish Pb-O-Pb random network.
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.
Shimazaki, Yosuke; Yamazaki, Kazunori; Iigaki, Kazuhiko
Hozengaku, 18(1), p.16 - 20, 2019/04
no abstracts in English