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Journal Articles

Reduction of the source term of an assumed criticality accident in a fuel fabrication facility with solution system

Fukaya, Yuji; Goto, Minoru

Annals of Nuclear Energy, 164, p.108617_1 - 108617_6, 2021/12

A reasonable source term of a hypothetical criticality accident for fuel fabrication facility with solution system has been proposed. The public exposure must not exceed the limitation of 5 mSv during an accident. Then, we proposed the reasonable source term of the first burst peak due to the hydrogen gas generation by radiation decomposition of water. With the criticality control system composed of the Criticality Accident Alarm System (CAAS) and soluble neutron absorber, safety is ensured by the reduced fission number. We confirmed the effect by environmental impact assessment during a criticality accident by using site condition of a fuel fabrication facility in Tokai-mura, Japan. As a result, the public exposure is reduced at a site boundary from 68 mSv to 0.6 mSv under the current regulatory guideline.

Journal Articles

Behavior of light elements in iron-silicate-water-sulfur system during early Earth's evolution

Iizuka, Riko*; Goto, Hirotada*; Shito, Chikara*; Fukuyama, Ko*; Mori, Yuichiro*; Hattori, Takanori; Sano, Asami; Funakoshi, Kenichi*; Kagi, Hiroyuki*

Scientific Reports (Internet), 11(1), p.12632_1 - 12632_10, 2021/06

 Times Cited Count:0 Percentile:0.01(Multidisciplinary Sciences)

The Earth's core consist of Fe-Ni alloy with some light elements (H, C, O, Si, S etc.). Hydrogen (H) is the most abundant element in the universe and one of the promising candidates. In this study, we have investigated the effects of sulfur(S) on hydrogenation of iron-hydrous silicate system containing saturated water in the ideal composition of the primitive Earth. We observed a series of phase transitions of Fe, dehydration of the hydrous mineral, and formation of olivine and enstatite with increasing temperature. The FeS formed as the coexisting phase of Fe under high-pressure and temperature condition, but its unit cell volume did not increase, suggesting that FeS is hardly hydrogenated. Recovered samples exhibited that H and S can be incorporated into solid Fe, which lowers the melting temperature as Fe(H$$_{x}$$)-FeS system. No detection of other light elements (C, O, Si) in solid Fe suggests that they dissolve into molten iron hydride and/or FeS in the later process of Earth's core-mantle differentiation.

Journal Articles

Feasibility study on burnable poison credit concept to HTGR fuel fabrication from core specification perspective

Fukaya, Yuji; Ueta, Shohei; Goto, Minoru; Ohashi, Hirofumi

Annals of Nuclear Energy, 151, p.107937_1 - 107937_9, 2021/02

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

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.

Journal Articles

Reactor physics experiment in a graphite-moderation system for HTGR

Fukaya, Yuji; Goto, Minoru; Nakagawa, Shigeaki; Nakajima, Kunihiro*; Takahashi, Kazuki*; Sakon, Atsushi*; Sano, Tadafumi*; Hashimoto, Kengo*

EPJ Web of Conferences, 247, p.09017_1 - 09017_8, 2021/02

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.

Journal Articles

Reactor noise analysis for a graphite-moderated and -reflected core in KUCA

Sakon, Atsushi*; Nakajima, Kunihiro*; Takahashi, Kazuki*; Hohara, Shinya*; Sano, Tadafumi*; Fukaya, Yuji; Hashimoto, Kengo*

EPJ Web of Conferences, 247, p.09009_1 - 09009_8, 2021/02

In graphite-reflected thermal reactors, even a detector placed far from fuel region may detect a certain degree of the correlation amplitude. This is because mean free path of neutrons in graphite is longer than that in water or polyethylene. The objective of this study is experimentally to confirm a high flexibility of neutron detector placement in graphite reflector for reactor noise analysis. The present reactor noise analysis was carried out in a graphite-moderated and -reflected thermal core in Kyoto University Critical Assembly (KUCA). BF$$_{3}$$ proportional neutron counters (1" dia.) were placed in graphite reflector region, where the counters were separated by about 35cm and 30cm -thick graphite from the core, respectively. At a critical state and subcritical states, time-sequence signal data from these counters were acquired and analyzed by a fast Fourier transform (FFT) analyzer, to obtain power spectral density in frequency domain. The auto-power spectral density obtained from the counters far from the core contained a significant degree of correlated component. A least-squares fit of a familiar formula to the auto-power spectral density data was made to determine the prompt-neutron decay constant. The decay constant was 63.3$$pm$$14.5 [1/s] in critical state. The decay constant determined from the cross-power spectral density and coherence function data between the two counters also had a consistent value. It is confirmed that reactor noise analysis is possible using a detector placed at about 35cm far from the core, as we expected.

Journal Articles

Influences of the ZrC coating process and heat treatment on ZrC-coated kernels used as fuel in Pu-burner high temperature gas-cooled reactor in Japan

Aihara, Jun; Ueta, Shohei; Honda, Masaki*; Mizuta, Naoki; Goto, Minoru; Tachibana, Yukio; Okamoto, Koji*

Journal of Nuclear Science and Technology, 58(1), p.107 - 116, 2021/01

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

The concept of a Pu-burner high temperature gas-cooled reactor (HTGR) has been proposed for purpose of more safely reducing amount of recovered Pu. This concept employs coated fuel particles (CFPs) with ZrC coated PuO$$_{2}$$-YSZ kernel and with tristructural (TRISO) coating for very high Pu burn-up and high nuclear proliferation resistance. In this report, we investigate the microstructure of the region that includes the surface of an as-fabricated CeO$$_{2}$$-YSZ kernel simulating PuO$$_{2}$$-YSZ kernel. We found both Zr-rich grains and Ce-rich grains to be densely distributed in that region including surface of CeO$$_{2}$$-YSZ kernel. On the other hand, it has been reported that there was a porous region near surface of the CeO$$_{2}$$-YSZ kernel of Batch I. This finding confirms that Ce-rich grains near surface of CeO$$_{2}$$-YSZ kernels coated with ZrC layers have been corroded during the deposition of the ZrC layer, whereas the Zr-rich grains were hardly affected.

Journal Articles

Recent R&D of HTGR and requirement for nuclear data

Fukaya, Yuji; Goto, Minoru; Nakagawa, Shigeaki

JAEA-Conf 2020-001, p.27 - 32, 2020/12

Recently, HTGR attracts a particular attention due to the outstanding safety features especially after the Fukushima Daiichi nuclear disaster, and the R&D is significantly promoted. In this presentation, we introduce the R&D plan of HTGR and the activities related to reactor physics and nuclear data including an experiment by using KUCA. Furthermore, requirement for nuclear data from the HTGR design is discussed.

Journal Articles

Status of investigation to ensure applicability of ECCS acceptance criteria to high-burnup fuel

Ozawa, Masaaki*; Amaya, Masaki

Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(4), p.185 - 200, 2020/12

no abstracts in English

Journal Articles

Research and development activities of JAEA for HTGR system realization

Mineo, Hideaki; Nishihara, Tetsuo; Ohashi, Hirofumi; Goto, Minoru; Sato, Hiroyuki; Takegami, Hiroaki

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 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.

Journal Articles

Guidance for developing fuel design limit of high temperature gas-cooled reactor

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.

Journal Articles

Methodology development for transient flow distribution analysis in high temperature gas-cooled reactor

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.

Journal Articles

Development of a flow network calculation code (FNCC) for high temperature gas-cooled reactors (HTGRs)

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.

Journal Articles

Uncertainty analysis of toxic gas leakage accident in cogeneration high temperature gas-cooled reactor

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.

Journal Articles

Crystal and magnetic structures of double hexagonal close-packed iron deuteride

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

 Times Cited Count:0 Percentile:0(Multidisciplinary Sciences)

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 $$varepsilon$$' 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$$_{0.68(1)}$$ at 673 K and 6.1 GPa and FeD$$_{0.74(1)}$$ 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$$_{1.0}$$. In the dhcp FeD$$_{1.0}$$ 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 $$pm$$ 0.06 B/Fe-atom aligned ferromagnetically in parallel on the Fe planes.

Journal Articles

Self-shielding effect of double heterogeneity for plutonium burner HTGR design

Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi

Annals of Nuclear Energy, 138, p.107182_1 - 107182_9, 2020/04

 Times Cited Count:0 Percentile:0.01(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 $$^{240}$$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 $$^{239}$$Pu and $$^{240}$$Pu in EOL.

Journal Articles

Research and development for safety and licensing of HTGR cogeneration system

Sato, Hiroyuki; Aoki, Takeshi; Ohashi, Hirofumi; Yan, X.

Nuclear Engineering and Design, 360, p.110493_1 - 110493_8, 2020/04

 Times Cited Count:2 Percentile:66.08(Nuclear Science & Technology)

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.

JAEA Reports

Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR) (FY2018)

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$$^{circ}$$C temperature of the outlet coolant with 30 MW 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.

Journal Articles

Development of numerical analysis code LEAP-III for tube failure propagation

Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki

Nihon 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.

Journal Articles

Reactor physics experiment in a graphite-moderation system for HTGR

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.

Journal Articles

Strong flux pinning by columnar defects with directionally dependent morphologies in GdBCO-coated conductors irradiated with 80 MeV Xe ions

Sueyoshi, Tetsuro*; Kotaki, Tetsuya*; Furuki, Yuichi*; Fujiyoshi, Takanori*; Semboshi, Satoshi*; Ozaki, Toshinori*; Sakane, Hitoshi*; Kudo, Masaki*; Yasuda, Kazuhiro*; Ishikawa, Norito

Japanese Journal of Applied Physics, 59(2), p.023001_1 - 023001_7, 2020/02

 Times Cited Count:3 Percentile:62.13(Physics, Applied)

We show that Xe ion irradiation with 80 MeV to GdBa$$_{2}$$Cu$$_{3}$$Oy-coated conductors creates different morphologies of columnar defects (CDs) depending on the irradiation angles relative to the c-axis: continuous CDs with a larger diameter are formed for oblique irradiation at $$theta_{rm i}$$ = 45$$^{circ}$$, whereas the same ion beam at a different angle ($$theta_{rm i}$$ = 0$$^{circ}$$) induces the formation of discontinuous CDs. The direction-dependent morphologies of CDs significantly affect the angular behavior of the critical current density $$J_{rm c}$$.

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