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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.

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

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

Experimental study on heat removal performance of a new Reactor Cavity Cooling System (RCCS)

Hosomi, Seisuke*; Akashi, Tomoyasu*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Takamatsu, Kuniyoshi

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 7 Pages, 2018/11

A new RCCS with passive safety features consists of two continuous closed regions. One is a region surrounding RPV. The other is a cooling region with heat transferred to the ambient air. The new RCCS needs no electrical or mechanical driving devices. We started experiment research with using a scaled-down test section. Three experimental cases under different emissivity conditions were performed. We used Monte Carlo method to evaluate the contribution of radiation to the total heat released from the heater. As a result, after the heater wall was painted black, the contribution of radiation to the total heat could be increased to about 60%. A high emissivity of RPV surface is very effective to remove more heat from the reactor. A high emissivity of the cooling part wall is also effective because it not only increases the radiation emitted to the ambient air, but also may increase the temperature difference among the walls and enhance the convection heat transfer in the RCCS.

Journal Articles

Feasibility study of large-scale production of iodine-125 at the high temperature engineering test reactor

Ho, H. Q.; Honda, Yuki*; Hamamoto, Shimpei; Ishii, Toshiaki; Fujimoto, Nozomu*; Ishitsuka, Etsuo

Applied Radiation and Isotopes, 140, p.209 - 214, 2018/10

 Percentile:100(Chemistry, Inorganic & Nuclear)

Journal Articles

Feasibility study of new applications at the high-temperature gas-cooled reactor

Ho, H. Q.; Honda, Yuki*; Hamamoto, Shimpei; Ishii, Toshiaki; Takada, Shoji; Fujimoto, Nozomu*; Ishitsuka, Etsuo

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10

Journal Articles

Conceptual study of an experimental HTGR upgraded from HTTR

Goto, Minoru; Fukaya, Yuji; Mizuta, Naoki; Inaba, Yoshitomo; Ohashi, Hirofumi; Yan, X.

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10

The HTTR (High Temperature engineering Test Reactor) constructed at JAEA-Oarai R&D center is a block-type experimental HTGR (High Temperature Gas-cooled Reactor) with 30 MW thermal power. It attained the first criticality at 1998 and has yielded very useful data for future HTGR design. Although the HTTR was designed very conservatively because the HTTR is the first HTGR for Japan, future HTGRs can be designed with a reasonable conservativeness based on the HTTR data. Additionally, it is possible to enhance the performance of the reactor core by improving the design and introducing new technologies. This paper describes a concept of an experimental HTGR that is upgraded from the HTTR by the reasonable conservativeness, the design improvement and the new technology introduction.

Journal Articles

Numerical evaluation on fluctuation absorption characteristics based on nuclear heat supply fluctuation test using HTTR

Takada, Shoji; Honda, Uki*; Inaba, Yoshitomo; Sekita, Kenji; Nemoto, Takahiro; Tochio, Daisuke; Ishii, Toshiaki; Sato, Hiroyuki; Nakagawa, Shigeaki; Sawa, Kazuhiro*

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10

Nuclear heat utilization systems connected to HTGRs will be designed on the basis of non-nuclear grade standards for easy entry of chemical plant companies, requiring reactor operations to continue even if abnormal events occur in the systems. The inventory control is considered as one of candidate methods to control reactor power for load following operation for siting close to demand area, in which the primary gas pressure is varied while keeping the reactor inlet and outlet coolant temperatures constant. Numerical investigation was carried out based on the results of nuclear heat supply fluctuation tests using HTTR by non-nuclear heating operation to focus on the temperature transient of the reactor core bottom structure by imposing stepwise fluctuation on the reactor inlet temperature under different primary gas pressures below 120C. As a result, it was emerged that the fluctuation absorption characteristics are not deteriorated by lowering pressure. It was also emerged that the reactor outlet temperature did not reach the scram level by increasing the reactor inlet temperature 10 C stepwise at 80% of the rated power as same with the full power case.

JAEA Reports

Excellent feature of Japanese HTGR technologies

Nishihara, Tetsuo; Yan, X.; Tachibana, Yukio; Shibata, Taiju; Ohashi, Hirofumi; Kubo, Shinji; Inaba, Yoshitomo; Nakagawa, Shigeaki; Goto, Minoru; Ueta, Shohei; et al.

JAEA-Technology 2018-004, 182 Pages, 2018/07

JAEA-Technology-2018-004.pdf:18.14MB

Research and development on High Temperature Gas-cooled Reactor (HTGR) in Japan started since late 1960s. Japan Atomic Energy Agency (JAEA) in cooperation with Japanese industries has researched and developed system design, fuel, graphite, metallic material, reactor engineering, high temperature components, high temperature irradiation and post irradiation test of fuel and graphite, high temperature heat application and so on. Construction of the first Japanese HTGR, High Temperature engineering Test Reactor (HTTR), started in 1990. HTTR achieved first criticality in 1998. After that, various test operations have been carried out to establish the Japanese HTGR technologies and to verify the inherent safety features of HTGR. This report presents several system design of HTGR, the world-highest-level Japanese HTGR technologies, JAEA's knowledge obtained from construction, operation and management of HTTR and heat application technologies for HTGR.

Journal Articles

Proposal of a neutron transmutation doping facility for n-type spherical silicon solar cell at high-temperature engineering test reactor

Ho, H. Q.; Honda, Yuki; Motoyama, Mizuki*; Hamamoto, Shimpei; Ishii, Toshiaki; Ishitsuka, Etsuo

Applied Radiation and Isotopes, 135, p.12 - 18, 2018/05

 Times Cited Count:1 Percentile:38.14(Chemistry, Inorganic & Nuclear)

Journal Articles

The Development status of Generation IV reactor systems, 2; High temperature gas-cooled reactor (HTGR)

Kunitomi, Kazuhiko; Nishihara, Tetsuo; Yan, X.; Tachibana, Yukio; Shibata, Taiju

Nippon Genshiryoku Gakkai-Shi, 60(4), p.236 - 240, 2018/04

High temperature gas-cooled reactor (HTGR) is a graphite-moderated and helium-gas-cooled thermal-neutron reactor that has excellent safety features and can produce high temperature heat of 950$$^{circ}$$C. It is expected to use for various heat applications as well as for electricity generation to reduce carbon dioxide emission. Japan Atomic Energy Agency (JAEA) has been promoted research and development to demonstrate the HTGR safety features using High temperature engineering test reactor (HTTR) and it's heat application. JAEA are also conducting the action to international deployment of Japanese HTGR technologies in cooperation with industries-government-academia. This paper reports status of the research and development of HTGR and domestic and international collaborations.

Journal Articles

HTTR-GT/H$$_{2}$$ test plant; System performance evaluation for HTTR gas turbine cogeneration plant

Sato, Hiroyuki; Nomoto, Yasunobu*; Horii, Shoichi*; Sumita, Junya; Yan, X.

Nuclear Engineering and Design, 329, p.247 - 254, 2018/04

 Percentile:100(Nuclear Science & Technology)

This paper presents the system performance evaluation for HTTR gas turbine cogeneration test plant (HTTR-GT/H$$_{2}$$ plant) so as to confirm that the design meets the requirements with respect to the demonstration test objective. Start-up and shut down operation sequences as well as operability of load following operation were investigated. In addition, system dynamic and control analyses for the test plant in the events of loss of generator load and upset of H$$_{2}$$ plant were performed. The simulation results presented in the paper show that the test plant is suitable for the test bed to validate control schemes against postulated transients in the GTHTR300C. The results also lead us to the conclusion that HTTR-GT/H$$_{2}$$ plant can be used to test operational procedures unique to HTGR direct-cycle gas turbine cogeneration.

Journal Articles

Design of HTTR-GT/H$$_{2}$$ test plant

Yan, X.; Sato, Hiroyuki; Sumita, Junya; Nomoto, Yasunobu*; Horii, Shoichi*; Imai, Yoshiyuki; Kasahara, Seiji; Suzuki, Koichi*; Iwatsuki, Jin; Terada, Atsuhiko; et al.

Nuclear Engineering and Design, 329, p.223 - 233, 2018/04

 Times Cited Count:2 Percentile:16.17(Nuclear Science & Technology)

The pre-licensing design of an HTGR cogeneration test plant to be coupled to JAEA's existing test reactor HTTR is presented. The plant is designed to demonstrate the system of JAEA commercial plant design GTHTR300C. With construction planned to be completed around 2025, the test plant is expected to be the first-of-a-kind nuclear system operating on two of the advanced energy conversion systems attractive for the HTGR closed cycle helium gas turbine for power generation and thermochemical iodine-sulfur water-splitting process for hydrogen production.

Journal Articles

3D crash calculation of stack and reactor building of HTTR

Ono, Masato; Fujiwara, Yusuke; Iigaki, Kazuhiko; Matsumoto, Tetsuro*; Taki, Nobuhiro*

Proceedings of European Research Reactor Conference 2018 (RRFM 2018) (Internet), 7 Pages, 2018/03

Integrity confirmation of building against the crash of flying object due to the tornado was carried out by formulas which calculate by simple shape. However, there was no study on crash calculation using complex shape such as the stack and reactor building. In this study, the crash calculation was carried out by a real shape model of High Temperature Engineering Test Reactor (HTTR) stack and reactor building using three-dimensional analysis code (VPS: Virtual Performance Solution). In the calculation, parameters of VPS were conservatively set in accordance with the formulas, which are formulated based on results of crash experiments and approved by the Nuclear Regulatory Authority. The crash calculation of stack and reactor building of HTTR was carried out using VPS. As a result, the integrity of building against the crash by stack was confirmed.

Journal Articles

Investigation of uncertainty caused by random arrangement of coated fuel particles in HTTR criticality calculations

Ho, H. Q.; Honda, Yuki; Goto, Minoru; Takada, Shoji

Annals of Nuclear Energy, 112, p.42 - 47, 2018/02

 Times Cited Count:1 Percentile:100(Nuclear Science & Technology)

Journal Articles

Loss of core cooling test with one cooling line inactive in Vessel Cooling System of High-Temperature Engineering Test Reactor

Fujiwara, Yusuke; Nemoto, Takahiro; Tochio, Daisuke; Shinohara, Masanori; Ono, Masato; Takada, Shoji

Journal of Nuclear Engineering and Radiation Science, 3(4), p.041013_1 - 041013_8, 2017/10

In HTTR, the test was carried out at the reactor thermal power of 9 MW under the condition that one cooling line of VCS was stopped to simulate the partial loss of cooling function from the surface of RPV in addition to the loss of forced cooling flow in the core simulation. The test results showed that temperature change of the core internal structures and the biological shielding concrete was slow during the test. Temperature of RPV decreased several degrees during the test. The temperature decrease of biological shielding made of concrete was within 1$$^{circ}$$C. The numerical result simulating the detail configuration of the cooling tubes of VCS showed that the temperature rise of cooling tubes of VCS was about 15$$^{circ}$$C, which is sufficiently small, which did not significantly affect the temperature of biological shielding concrete. As the results, it was confirmed that the cooling ability of VCS can be kept in case that one cooling line of VCS is lost.

JAEA Reports

Design database of helium gas turbine for HTTR-GT/H$$_{2}$$ test plant (Revised version)

Imai, Yoshiyuki; Sato, Hiroyuki; Yan, X.

JAEA-Data/Code 2017-011, 39 Pages, 2017/08

JAEA-Data-Code-2017-011.pdf:2.93MB

This report is the revised version of the report titled "Design Database of Helium Gas Turbine for High Temperature Gas-cooled Reactor, JAEA-Data/Code 2016-007" reflecting component design and experimental data analysis results for fission product isotope diffusion through the turbine blade alloy conducted in Fiscal Year 2016.

Journal Articles

Thermal-hydraulic analyses of the High-Temperature engineering Test Reactor for loss of forced cooling at 30% reactor power

Takamatsu, Kuniyoshi

Annals of Nuclear Energy, 106, p.71 - 83, 2017/08

The HTTR, which is the only HTGR having inherent safety features in Japan, conducted a safety demonstration test involving a loss of both reactor reactivity control and core cooling. The paper shows thermal-hydraulics during the LOFC test at an initial power of 30% reactor power (9 MW), when the insertion of all control rods was disabled and all gas circulators were tripped to reduce the coolant flow rate to zero. The analytical results could show that the downstream of forced convection caused by the HPS pushes down the upstream by natural convection in the fuel assemblies; however, the forced convection has little influence on the core thermal-hydraulics without the reactor outlet coolant temperature. As a result, the three-dimensional thermal-phenomena inside the RPV during the LOFC test could be understood qualitatively.

Journal Articles

Determination of reactivity and neutron flux using modified neural network for HTGR

Subekti, M.*; Kudo, Kazuhiko*; Nabeshima, Kunihiko; Takamatsu, Kuniyoshi

Atom Indonesia, 43(2), p.93 - 102, 2017/08

Reactor kinetics based on point kinetic model have been generally applied as the standard method for neutronics codes. As the central control rod (C-CR) withdrawal test has demonstrated in a prismatic core of HTTR, the transient calculation of kinetic parameter, such as reactivity and neutron fluxes, requires a new method to shorten calculation-process time. Development of neural network method was applied to point kinetic model as the necessity of real-time calculation that could work in parallel with the digital reactivity meter. The combination of TDNN and Jordan RNN, such as TD-Jordan RNN, was the result of the modeling approach. The application of TD-Jordan RNN with adequate learning, tested offline, determined results accurately even when signal inputs were noisy. Furthermore, the preprocessing for neural network input utilized noise reduction as one of the equations to transform two of twelve time-delayed inputs into power corrected inputs.

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