Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 134

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Fuel cycle scenarios and back-end technologies of HTGR in Japan

Fukaya, Yuji; Goto, Minoru; Shibata, Taiju

IAEA-TECDOC-2040, p.133 - 136, 2023/12

Japan has developed back-end technologies to establish a multi-recycling fuel cycle with fast breeder reactors (FBRs) to ensure energy resources. Even though the development of FBR has been retreated to one of fundamental research, the reprocessing technologies for uranium fuel and disposal technologies had been completed for Light Water Reactor (LWR) fuel cycle on the process. These technologies were inherited to utilities and are about to be practical. Now, Japan had been completed High Temperature Engineering Test Reactor (HTTR) a prototype and research reactor, a commercial High Temperature Gas-cooled Reactor (HTGR) design Gas Turbine High Temperature Reactor 300 (GTHTR300) with related reprocessing technologies, and is planning domestic demonstration reactor project. In this context, a representative fuel cycle policy is reprocessing in Japan. However, Japan has investigated various fuel cycle scenarios to expand the usage of the commercial HTGR. Then, we would like to introduce the scenarios and development status of related technologies in the present study.

Journal Articles

Reactor physics experiment on a graphite-moderated core to construct integral experiment database for HTGR

Okita, Shoichiro; Fukaya, Yuji; Sakon, Atsushi*; Sano, Tadafumi*; Takahashi, Yoshiyuki*; Unesaki, Hironobu*

Nuclear Science and Engineering, 197(8), p.2251 - 2257, 2023/08

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

JAEA Reports

Study on disposal of waste from reprocessing for commercial HTGR spent fuel

Fukaya, Yuji; Maruyama, Takahiro; Goto, Minoru; Ohashi, Hirofumi; Higuchi, Hideaki

JAEA-Research 2023-002, 19 Pages, 2023/06

JAEA-Research-2023-002.pdf:1.48MB

A study on disposal of waste derived from commercial High Temperature Gas-cooled Reactor ("HTGR") has been performed. Because of significant difference between the reprocessing of Light Water Reactor ("LWR") and that of HTGR due to difference in structures of the fuel, adoptability of the laws relating to reprocessing waste disposal, which is enacted for LWR, to HTGR waste should be confirmed. Then, we compared the technologies and waste of reprocessing and evaluated radioactivity concentration in graphite waste by activation and contamination based on whole core burn-up calculation. As a result, it was found that SiC residue waste should be disposed of into a geological repository as 2nd class designated radioactive waste in the Designated Radioactive Waste Final Disposal Act (Act No.117 of 2000), by way of amendment of the applicable order, same as hull and end-piece of LWR, and graphite waste should be shallowly disposed of than geological disposal as 2nd class waste for pit disposal in the Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors (Act No.166 of 1957) same as a channel box of LWR.

Journal Articles

Feasibility study on reprocessing of HTGR spent fuel by existing PUREX plant and technology

Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi

Annals of Nuclear Energy, 181, p.109534_1 - 109534_10, 2023/02

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

Feasibility of reprocessing of High Temperature Gas-cooled Reactor (HTGR) spent fuel by existing Plutonium Uranium Redox EXtraction (PUREX) plant and technology has been investigated. The spent fuel dissolved solution includes approximately 3 times amount of uranium-235 and 1.5 times amount of protonium because of the 3 times higher burnup compared with that of Light Water Reactor (LWR). Then, the heavy metal of the spent fuel is planned to be diluted to 3.1 times by depleted uranium to satisfy the limitation of Rokkasho Reprocessing Plant (RRP) plant. In the present study, recoverability of uranium and plutonium with the dilution is confirmed by a simulation with a reprocessing process calculation code. Moreover, the case without the dilution from the economic perspective is investigated. As a result, the feasibility is confirmed without the dilution, and it is expected that the reprocessed amount is reduced to 1/3 compared with a diluted case even though the facility should be optimized from the perspective of mass flow and criticality.

Journal Articles

Study on evaluation method of kernel migration of TRISO fuel for High Temperature Gas-cooled Reactor

Fukaya, Yuji; Okita, Shoichiro; Sasaki, Koei; Ueta, Shohei; Goto, Minoru; Ohashi, Hirofumi; Yan, X.

Nuclear Engineering and Design, 399, p.112033_1 - 112033_9, 2022/12

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

Kernel migration of TRi-structural ISOtropic (TRISO) fuel for High Temperature Gas-cooled Reactor (HTGR) has been analyzed to investigate the potential dominating effects. Kernel migration is a major fuel failure mode and dominant to determine the lifetime of the fuel for High Temperature engineering Test Reactor (HTTR). However, this study shows that the result and reliability depend on the evaluation method. The evaluation method used in this study takes into account of actual distribution of Coated Fuel Particles (CFPs) and the resulting heterogeneous fuel temperature calculation with such distribution. The result shows that the Kernel Migration Rate (KMR) is predicted to be about 10% less compared with the most conservative evaluation.

Journal Articles

Reactor physics experiment in a graphite moderation system for HTGR, 3

Fukaya, Yuji; Okita, Shoichiro; Kanda, Shun*; Goto, Masaki*; Nakajima, Kunihiro*; Sakon, Atsushi*; Sano, Tadafumi*; Hashimoto, Kengo*; Takahashi, Yoshiyuki*; Unesaki, Hironobu*

KURNS Progress Report 2021, P. 101, 2022/07

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) in 2018. The objectives are to intro-duce the generalized bias factor method to avoid full mock-up experiment for the first commercial HTGR and to improve neutron instrumentation system by virtue of the particular characteristics due to a graphite moderation system. For this end, we composed B7/4"G2/8"p8EU(3)+3/8"p38EU in the B-rack of Kyoto University Critical Assembly (KUCA) in 2021.

Journal Articles

Reactor noise power-spectral analysis for a graphite-moderated and -reflected core, 3

Sakon, Atsushi*; Hashimoto, Kengo*; Sano, Tadafumi*; Nakajima, Kunihiro*; Kanda, Shun*; Goto, Masaki*; Fukaya, Yuji; Okita, Shoichiro; Fujimoto, Nozomu*; Takahashi, Yoshiyuki*

KURNS Progress Report 2021, P. 100, 2022/07

The R&D of reactor noise analysis to obtain HTGR nuclear characteristics have been performed with Kyoto University Critical Assembly (KUCA). In the last study, a neutron detector located about 55 cm away of fuel assembly measured the auto power spectral density. However, the prompt neutron decay constants obtained by this detector was different from that of other detectors. The objective of this study is experimental study of reactor noise analysis by the power spectrum method using neutron detector placed outside reactor core.

Journal Articles

Re-evaluation of electricity generation cost of HTGR

Fukaya, Yuji; Ohashi, Hirofumi; Sato, Hiroyuki; Goto, Minoru; Kunitomi, Kazuhiko

Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 21(2), p.116 - 126, 2022/06

An improvement electricity generation cost evaluation method for High Temperature Gas-cooled Reactors (HTGRs) has been performed. Japan Atomic Energy Agency (JAEA) had completed the commercial HTGR concept named Gas Turbine High Temperature Reactor (GTHTR300) and the electricity generation cost evaluation method approximately a decade ago. The cost evaluation was developed based on the method of Federation of Electric Power Companies (FEPC). The FEPC method was drastically revised after the Fukushima Daiichi nuclear disaster. Moreover, the escalation of material and labor cost for the decade should be consider to evaluate the latest cost. Therefore, we revised the cost evaluation method for GTHTR300 and the cost was compared with that of Light Water Reactor (LWR). As a result, it was found that the electricity generation cost of HTGR of 7.9 yen/kWh is cheaper than that of LWR of 11.7 yen/kWh by approximately 30% at the capacity factor of 70%.

Journal Articles

Preliminary experiment in a graphite-moderated core to avoid full mock-up experiment for the future first commercial HTGR

Okita, Shoichiro; Fukaya, Yuji; Sakon, Atsushi*; Sano, Tadafumi*; Takahashi, Yoshiyuki*; Unesaki, Hironobu*

Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 9 Pages, 2022/05

Journal Articles

Computed tomography neutron detector system to observe power distribution in a core with long neutron flight path

Fukaya, Yuji; Okita, Shoichiro; Nakagawa, Shigeaki; Goto, Minoru; Ohashi, Hirofumi

Annals of Nuclear Energy, 168, p.108911_1 - 108911_7, 2022/04

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

A power distribution monitoring system by using a moving detector for a core with a long neutron flight path has been proposed. High Temperature Gas-cooled Reactor (HTGR) and Fast Reactor (FR) has a long neutron flight path and the neutrons reach to detector far from fuel assembly in the center of the core unlike Light Water Reactor (LWR). By using the feature, power distribution can be observed with a few detectors by moving the detector and computed tomography technology similar to X-ray Computed Tomography (CT). For a small-sized core, the power distribution can be evaluated only by an ex-core neutron detector. For a large-sized core with inner detectors, the power distribution can be observed with a small number of in-core detectors even if the deployment is limited due to material integrity conditions such as temperature environment. The feasibility is numerically confirmed by simulations of the HTGR core and its detector response. It is expected to observe the power distribution in the core of HTGR and FR, which is difficult continuously to deploy in-core detectors because of high temperature and/or high irradiation damage.

Journal Articles

Toxicity reduction with total volume control in nuclear waste

Fukaya, Yuji; Ueta, Shohei; Yamamoto, Tomohiko; Chikazawa, Yoshitaka; Yan, X.

Nuclear Technology, 208(2), p.335 - 346, 2022/02

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

When the total volume control on toxicity for nuclear waste management is applied, it becomes a limiting factor for the permittable total operation capacity of nuclear reactors. An alternative conceptual scenario to achieve the control is proposed that aims at toxicity reduction through Partitioning and Transmutation (P&T). Specifically, the electricity generation capacity could be inversely increased up with transmutation of $$^{90}$$Sr-$$^{137}$$Cs. Simultaneously, the cooling time before disposal is reduced to 50 years from the 300 years required by the existing scenarios such as (Accelerator Driven System (ADS). Finally, the scenario is also found feasible in terms of energy balance and cost by the neutron source of Li(d,xn) reaction with the deuteron accelerator for transmutation.

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

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

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

A Pseudo-material method for graphite with arbitrary porosities in Monte Carlo criticality calculations

Okita, Shoichiro; Nagaya, Yasunobu; Fukaya, Yuji

Journal of Nuclear Science and Technology, 58(9), p.992 - 998, 2021/09

 Times Cited Count:2 Percentile:34.45(Nuclear Science & Technology)

Journal Articles

Manufacturability estimation on burnable poison mixed fuel for improving criticality safety of HTGR fuel fabrication

Hasegawa, Toshinari; Fukaya, Yuji; Ueta, Shohei; Goto, Minoru

Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 5 Pages, 2021/08

Burnable poison (BP) credit concept has been proposed as a criticality safety measure for commercial high temperature gas-cooled reactor (HTGR) fuel fabrication, so we estimated manufacturability of the BP-mixed UO$$_2$$ kernel for the practical use of the concept. As a BP, boron, gadolinium, erbium, and hafnium are investigated. Boron mixed fuel kernels are fabricated by mixing boric acid powder with U$$_3$$O$$_8$$ powder. In the case of the other BPs, BP nitrate powder is mixed with U$$_3$$O$$_8$$ powder. In order to confirm that BP remain in the kernels after the heat treatment processes, thermodynamic equilibrium analysis was performed. Above 450$$^circ$$C, boron would melt and vaporize during the heat treatment processes, so it was found that the boron mixed fuel kernel fabrication is difficult. On the other hand, it was found that gadolinium, erbium, and hafnium would change to solid oxides that do not melt and vaporize even at 2000$$^circ$$C, and there was no problem with manufacturability of the BP-mixed fuel kernel.

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:2 Percentile:34.45(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

High temperature gas-cooled reactors

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 950$$^{circ}$$C, 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.

Journal Articles

Derivation of ideal power distribution to minimize the maximum kernel migration rate for nuclear design of pin-in-block type HTGR

Okita, Shoichiro; Fukaya, Yuji; Goto, Minoru

Journal of Nuclear Science and Technology, 58(1), p.9 - 16, 2021/01

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

Suppressing the kernel migration rates, which depend on both the fuel temperature and the fuel temperature gradient, under normal operation condition is quite important from the viewpoint of the fuel integrity for High Temperature Gas-cooled Reactors. The presence of the ideal axial power distribution to minimize the maximum kernel migration rate allows us to improve efficiency of design work. Therefore, we propose a new method based on Lagrange multiplier method in consideration of thermohydraulic design in order to obtain the ideal axial power distribution to minimize the maximum kernel migration rate. For one of the existing conceptual designs performed by JAEA, the maximum kernel migration rate for the power distribution to minimize the maximum kernel migration rate proposed in this study is lower by approximately 10% than that for the power distribution as a conventional design target to minimize the maximum fuel temperature.

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.

134 (Records 1-20 displayed on this page)