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Journal Articles

Comparative methodology between actual RCCS and downscaled heat-removal test facility

Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*

Annals of Nuclear Energy, 133, p.830 - 836, 2019/11

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.

JAEA Reports

Estimation of exchange time for neutron startup sources of HTTR

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 $$^{252}$$Cf to start up the reactor and to confirm count rates of wide range monitors. The half-life of $$^{252}$$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 $$^{252}$$Cf by 2024.

Journal Articles

Reactor physics experiment in graphite moderation system for HTGR, 1

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.

JAEA Reports

Flow separation at inlet causing transition and intermittency in circular pipe flow

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.

JAEA Reports

Report of summer holiday practical training 2018; Feasibility study on nuclear battery using HTTR core; feasibility study for nuclear design

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.

JAEA Reports

Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR); FY2017

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.

Journal Articles

Study on levelizing electricity generation cost for nuclear power generation between generations

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.

Journal Articles

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

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$$_{2}$$ 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$$_{2}$$ 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.

Journal Articles

Establishment of numerical model to investigate heat transfer and flow characteristics by using scale model of vessel cooling system for HTTR

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 200$$^{circ}$$C 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-$$omega$$-SST model was selected for turbulent calculation attributed to predict the effects mentioned above adequately. The numerical results using the k-$$omega$$-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-$$varepsilon$$ model which does not account the effects.

Journal Articles

Development of container using plasma sprayed and laser treated material for sulfuric acid decomposition of thermochemical water-splitting iodine-sulfur process

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$$_{2}$$ 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.

Journal Articles

Research and development on high burnup HTGR fuels in JAEA

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 $$^{circ}$$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.

Journal Articles

Study of SiC-matrix fuel element for HTGR

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.

Journal Articles

Module design of silica membrane reactor for hydrogen production via thermochemical IS process

Odtsetseg, M.; Tanaka, Nobuyuki; Nomura, Mikihiro*; Kubo, Shinji

International Journal of Hydrogen Energy, 44(21), p.10207 - 10217, 2019/04

Journal Articles

Development of hydrogen production technology by thermal water splitting with solar heat

Inagaki, Yoshiyuki; Sakaba, Nariaki

Shokubai, 61(2), p.92 - 96, 2019/04

The outline of the membrane IS process to produce hydrogen by thermochemical water splitting using solar heat at around 650$$^{circ}$$C is described. The membrane technology has been applied to the three main reaction of the IS process to lower the reaction temperature and reduce the amount of circulation materials in the process. The key component technologies such as catalysts, membranes and corrosion resistant materials have been developed. The study was supported in part by the Council for Science, Technology and Innovation, Cross-ministerial Strategic Innovation Promotion Program, "Energy Carrier".

Journal Articles

Development of fabrication technology for oxidation-resistant fuel elements for high-temperature gas-cooled reactors

Aihara, Jun; Honda, Masaki*; Ueta, Shohei; Ogawa, Hiroaki; Ohira, Koichi*; Tachibana, Yukio

Nippon Genshiryoku Gakkai Wabun Rombunshi, 18(1), p.29 - 36, 2019/03

Japan Atomic Energy Agency carried out development of fabrication technology of oxidation resistant fuel element for improvement of safety of high temperature gas-cooled reactors in serious oxidation accident, based on precursor research in former JAEA. Dummy coated fuel particles (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 dummy oxidation resistant fuel elements with SiC/C mixed matrix. We fabricated dummy oxidation resistant fuel elements with matrix whose Si/C mole ratio (about 0.551) is three times as large as that in precursor research. Si peak was not detected by X-ray diffraction of matrix. Better oxidation resistant was confirmed with oxidation test in 20% O$$_{2}$$ at 1673 K than that of ordinal fuel compact with ordinal graphite/carbon matrix. All dummy coated fuel particles were held in specimen after 10 h oxidation.

Journal Articles

Study of an HTGR and renewable energy hybrid system for grid stability

Sato, Hiroyuki; Yan, X.

Nuclear Engineering and Design, 343, p.178 - 186, 2019/03

 Percentile:100(Nuclear Science & Technology)

A hybrid system combining HTGR and renewable energy is investigated to compensate intermittent renewable energy power generation. A new proposal of using the inventory and bypass control devices already built in the gas turbine, is found to be effective to compensate hourly to daily variation of renewable energy. The reactor thermal power remains at constant full power while the heat output is increased or decreased subject to the need of reactor power generation. On the other hand, the massive heat capacity in the graphite core is shown to be sufficient to compensate renewable energy on a time scale of seconds to minutes and up to about 20% of the rated power output of the nuclear plant. Similarly, no additional control devices are required to perform this control operation. These findings demonstrate the technical and economic potential of the HTGR system to maintain the stability of a grid being incorporated with significant portfolios of renewable energy power generation.

Journal Articles

Improvement of heat-removal capability using heat conduction on a novel reactor cavity cooling system (RCCS) design with passive safety features through radiation and natural convection

Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*

Annals of Nuclear Energy, 122, p.201 - 206, 2018/12

 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. This study addresses an improvement of heat-removal capability using heat conduction on the RCCS. As a result, a heat flux removed by the RCCS could be doubled; therefore, it is possible to halve the height of the RCCS or increase the thermal reactor power.

JAEA Reports

Calculations of Tritium Recoil Release from Li and U Impurities in Neutron Reflectors (Joint research)

Ishitsuka, Etsuo; Kenzhina, I.*; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*

JAEA-Technology 2018-010, 33 Pages, 2018/11

JAEA-Technology-2018-010.pdf:2.58MB

As a part of study on the mechanism of tritium release to the primary coolant in research and testing reactors, tritium recoil release rate from Li and U impurities in the neutron reflector made by beryllium, aluminum and graphite were calculated by PHITS code. On the other hand, the tritium production from Li and U impurities in beryllium neutron reflectors for JMTR and JRR-3M were calculated by MCNP6 and ORIGEN2 code. By using both results, the amount of recoiled tritium from beryllium neutron reflectors were estimated. It is clear that the amount of recoiled tritium from Li and U impurities in beryllium neutron reflectors are negligible, and 2 and 5 orders smaller than that from beryllium itself, respectively.

Journal Articles

Nuclear and thermal feasibility of lithium-loaded high temperature gas-cooled reactor for tritium production for fusion reactors

Goto, Minoru; Okumura, Keisuke; Nakagawa, Shigeaki; Inaba, Yoshitomo; Matsuura, Hideaki*; Nakaya, Hiroyuki*; Katayama, Kazunari*

Fusion Engineering and Design, 136(Part.A), p.357 - 361, 2018/11

A High Temperature Gas-cooled Reactor (HTGR) is proposed as a tritium production device, which has the potential to produce a large amount of tritium using $$^{6}$$Li(n,$$alpha$$)T reaction. In the HTGR design, generally, boron is loaded into the core as a burnable poison to suppress excess reactivity. In this study, lithium is loaded into the HTGR core instead of boron and is used as a burnable poison aiming to produce thermal energy and tritium simultaneously. The nuclear characteristics and the fuel temperature were calculated to confirm the feasibility of the lithium-loaded HTGR. It was shown that the calculation results satisfied the design requirements and hence the feasibility was confirmed for the lithium-loaded HTGR, which produce thermal energy and tritium.

Journal Articles

Uranium-based TRU multi-recycling with thermal neutron HTGR to reduce environmental burden and threat of nuclear proliferation

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

 Percentile:100(Nuclear Science & Technology)

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.

1611 (Records 1-20 displayed on this page)