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Journal Articles

Fusion reaction $$^{48}$$Ca+$$^{249}$$Bk leading to formation of the element Ts (Z=117)

Khuyagbaatar, J.*; Yakushev, A.*; D$"u$llmann, Ch. E.*; Ackermann, D.*; Andersson, L.-L.*; Asai, Masato; Block, M.*; Boll, R. A.*; Brand, H.*; Cox, D. M.*; et al.

Physical Review C, 99(5), p.054306_1 - 054306_16, 2019/05

We have performed an experiment to synthesize the element 117 (Ts) with the $$^{48}$$Ca+$$^{249}$$Bk fusion reaction. Four $$alpha$$-decay chains attributed to the element 117 were observed. Two of them were long decay chains which can be assigned to the one originating from the $$alpha$$ decay of $$^{294}$$Ts. The other two were short decay chains which are consistent with the one originating from the $$alpha$$ decay of $$^{293}$$Ts. We have compared the present results with the literature data, and found that our present results mostly confirmed the literature data, leading to the firm confirmation of the synthesis of the element 117.

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

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

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.

Journal Articles

Conceptual design of the steam reforming system for hydrogen production connected to HTTR

Iwatsuki, Jin; Ohashi, Hirofumi; Yan, X.

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

The HTTR is a 30MW, 950$$^{circ}$$C high temperature engineering test reactor built and operated on the site of the Oarai Research & Development Center of Japan Atomic Energy Agency (JAEA). In the framework of the HTTR project, JAEA has been conducting a research and development on the steam reforming system (CH$$_{4}$$ + H$$_{2}$$O = 3H$$_{2}$$ + CO). JAEA had constructed a mock-up test facility in 2002, and investigated transient behavior of the hydrogen production system and established system controllability. Based on the results and experience of above, the conceptual design of steam reforming system for hydrogen production connected to HTTR has been studied. The system condition was optimized considering the HTTR specification and the experience on the construction and the operation of the mock-up test facility. The hydrogen production system is heated with about 0.2MW transported from the HTTR to the hydrogen system via a helium loop. The system produces about 70 Nm$$^{3}$$/h hydrogen.

Journal Articles

Conceptual design study of a high performance commercial HTGR

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

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

Conceptual design study of a high performance commercial HTGR has been performed at target output of 165MWt. Requirements for the HTGR are small-sized vessel for transportation, durability of vessel to irradiation damage, fuel reloading scheme to shorten the duration of reloading, low pressure drop fuel element, a small number of fuel enrichments, and so on. To satisfy the requirement, we investigated the core configuration, shielding and reflector configuration, fuel reloading scheme. As a result, we completed the design with the vessel diameter of 4.5m, which can be transported by any means, such as, by load, rail, ship, and air plane, and high load factor over 90%.

Journal Articles

Conceptual plant system design study of an experimental HTGR upgraded from HTTR

Ohashi, Hirofumi; Goto, Minoru; Ueta, Shohei; Sato, Hiroyuki; Fukaya, Yuji; Kasahara, Seiji; Sasaki, Koei; Mizuta, Naoki; Yan, X.; Aoki, Takeshi*

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

Conceptual design study of an experimental HTGR is performed to upgrade the plant system from Japanese High Temperature engineering Test Reactor (HTTR) to a commercial HTGR. Safety systems of HTTR are upgraded to demonstrate the commercial HTGR concept, such as a passive reactor cavity cooling system, a confinement, etc. An intermediate heat exchanger (IHX) is replaced by a steam generator (SG) for a process heat supply to demonstrate the technology for a commercial use. This paper describes the conceptual design study results of the plant system of the experimental HTGR.

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

Research and development for safety and licensing of HTGR cogeneration system

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

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

Japan Atomic Energy Agency 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. This paper explains 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 is also presented.

Journal Articles

RELAP5 modeling of the HTTR-GT/H$$_{2}$$ secondary system and turbomachinery

Humrickhouse, P. W.*; Sato, Hiroyuki; Imai, Yoshiyuki; Sumita, Junya; Yan, X.

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

This work describes the development of a RELAP5-3D model of the HTTR-GT/H$$_{2}$$ plant secondary system. The RELAP5-3D model presently includes detailed models of several of the heat exchangers in the secondary system as well as the turbomachinery, which includes two compressors and two gas turbines connected to a common shaft and motor. The predictions of the model agreed well to design parameters in both sole power generation and hydrogen co-generation modes in most instances. Both the turbomachinery and heat exchanger models rely on extensive customization via RELAP5-3D control variables, and these implementations are outlined in detail. Potential improvements to the RELAP5-3D turbine model are discussed.

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

A Concept of intermediate heat exchanger for high-temperature gas reactor hydrogen and power cogeneration system

Hirota, Noriaki; Terada, Atsuhiko; Yan, X.; Tanaka, Kohei*; Otani, Akihito*

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 7 Pages, 2018/07

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

Conceptual design of the iodine-sulfur process flowsheet with more than 50% thermal efficiency for hydrogen production

Kasahara, Seiji; Imai, Yoshiyuki; Suzuki, Koichi*; Iwatsuki, Jin; Terada, Atsuhiko; Yan, X.

Nuclear Engineering and Design, 329, p.213 - 222, 2018/04

 Percentile:100(Nuclear Science & Technology)

A conceptual design of a practical large scale plant of the thermochemical water splitting iodine-sulfur (IS) process flowsheet was carried out as a heat application of JAEA's commercial high temperature gas cooled reactor GTHTR300C plant design. Innovative techniques proposed by JAEA were applied for improvement of hydrogen production thermal efficiency; depressurized flash concentration H$$_{2}$$SO$$_{4}$$ using waste heat from Bunsen reaction, prevention of H$$_{2}$$SO$$_{4}$$ vaporization from a distillation column by introduction of H$$_{2}$$SO$$_{4}$$ solution from a flash bottom, and I$$_{2}$$ condensation heat recovery in an HI distillation column. Hydrogen of about 31,900 Nm$$^{3}$$/h would be produced by 170 MW heat from the GTHTR300C. A thermal efficiency of 50.2% would be achievable with incorporation of the innovative techniques and high performance HI concentration and decomposition components and heat exchangers expected in future R&D.

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.

JAEA Reports

Development plan of austenitic Fe and Ni based alloys with improved corrosion resistance to sulfuric acid and HI fluids of industrial processes

Hirota, Noriaki; Iwatsuki, Jin; Imai, Yoshiyuki; Yan, X.

JAEA-Technology 2017-027, 19 Pages, 2017/12

JAEA-Technology-2017-027.pdf:2.08MB

In this study, austenitic Fe-based alloys and Ni based alloys was developed as candidate structural materials for equipment operated in sulfuric acid and hydrogen iodide (HI) environment, which exists in various industrial processes including iodine-sulfur (IS) hydrogen production process and geothermal power generation process. The objectives of the study are to achieve the corrosion resistance performance sufficient under the working condition of these processes and to overcome the practical scale-up difficulty of the ceramic (SiC) material that is presently used in the processes due to the manufacturing size limitation of the ceramic. The chemical composition development plan for the austenitic Fe-based alloys is threefold: reinforcement of matrix by addition of Cu and Ta, strength compensation of the surface film by addition of Si and Ti, and prevention of peeling of surface oxide by addition of rare earth elements. Because addition of Cu and Si is known to reduce the ductility of the material and thus manufacturability of the component, it is important to determine the allowable amount of each element to be added. On the other hand, the chemical composition development plan for the Ni based alloys is reinforcement of matrix by addition of Mo, W and Ta, strength compensation of the surface film by addition of Ti, and prevention of peeling of surface oxide by addition of rare earth elements. In particular, the addition of Mo and W to the Ni based alloy is expected to be effective in preventing dimensional deviation of structures from increasing during heating and cooling of process equipment. Various material specimens will be fabricated based on the above chemical composition development plans and tests on these specimens will then be carried out to confirm the corrosion resistance performance under the fluid conditions simulating each industrial processes.

JAEA Reports

System analysis for HTTR-GT/H$$_{2}$$ plant; Safety analysis of HTTR for coupling helium gas turbine and H$$_{2}$$ plant

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

JAEA-Technology 2017-020, 23 Pages, 2017/08

JAEA-Technology-2017-020.pdf:1.23MB

JAEA initiated a nuclear cogeneration demonstration project with helium gas turbine power generation and thermochemical hydrogen production utilizing the HTTR. This study carries out system analysis for the HTTR gas turbine hydrogen cogeneration test plant. The evaluation was conducted for the events newly identified corresponding to the coupling of helium gas turbine and hydrogen production plant to the HTTR. The results showed that loss of load event does not have impact on temperature of fuel and reactor coolant pressure boundary. In addition, reactor coolant pressure does not exceed the evaluation criteria. Furthermore, it was shown that reactor operation can be maintained against temperature transients induced by abnormal events in hydrogen production plant.

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.

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