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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

Research and development on high burnup HTGR fuels in JAEA

Ueta, Shohei; Mizuta, Naoki; Sasaki, Koei; Sakaba, Nariaki; Ohashi, Hirofumi; Yan, X. L.

Mechanical Engineering Journal (Internet), 7(3), p.19-00571_1 - 19-00571_12, 2020/06

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/t. Furthermore, the future R&Ds needed for the high burnup HTGR fuel are described based on these experimental results.

Journal Articles

Conceptual design study of a high performance commercial HTGR for early introduction

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

Nuclear Engineering and Design, 361, p.110577_1 - 110577_6, 2020/05

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Conceptual design study of a commercial High Temperature Gas-cooled Reactor (HTGR) for early introduction has been performed based on the cumulated experience in design, construction, and operation of the High Temperature engineering Test Reactor (HTTR) and design of the commercial Gas Turbine High Temperature Reactor 300 (GTHTR300). The power output is 165 MWt and the inlet and outlet coolant temperatures are 325$$^{circ}$$C and 750$$^{circ}$$C, respectively, to provide steam for industrial utilization. However, given a requirement for the reactor pressure vessel to be smaller even that of the 30 MWt HTTR, several challenging technical problems have to be dealt with to arrive in a high performance core design that provides extended fuel burnup, prolonged refueling period, improved fuel refueling scheme, improved fuel element and so on from the HTTR.

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:100(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. L.

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

 Times Cited Count:0 Percentile:100(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.

Journal Articles

Development of new corrosion test equipment simulating sulfuric acid decomposition gas environment in a thermochemical hydrogen production process

Hirota, Noriaki; Kasahara, Seiji; Iwatsuki, Jin; Imai, Yoshiyuki; Ohashi, Hirofumi; Yan, X. L.; Tachibana, Yukio

Zairyo To Kankyo, 68(6), p.137 - 142, 2019/06

New corrosion test equipment for high temperature gas of decomposed sulfuric acid was manufactured in order to ascertain flow rate of sulfuric acid in the piping, occurrence of sulfuric acid decomposition reaction in the equipment, and temperature distribution inside the furnace tube. The flow rate of the sulfuric acid solution was constantly measured using an ultrasonic flowmeter. The SO$$_{3}$$ concentration at the inlet of the test equipment was almost the same as that at the inlet of the sulfuric acid decomposer in the hydrogen production plant assuming a high-temperature gas cooled reactor hydrogen-power cogeneration system (GTHTR300C). On the other hand, during a test, leakage of sulfuric acid occurred from the fitting part at the outlet side. Hence the temperature distribution of the fitting part at the outlet side was investigated using fluid analysis. As a result, it was found that the temperature at the fitting was low enough to use fluorine joint grease when the distance was 0.05 m or more away from the outlet side pipe. An improved furnace tube was manufactured and the temperature was measured again at fitting part of the outlet side. The temperature was lower that the temperature limit of the joint grease and almost the same as the temperature distribution in the fluid analysis, and leakage of sulfuric acid has not occurred to date.

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

Research and development on high burnup HTGR fuels in JAEA

Ueta, Shohei; Mizuta, Naoki; Sasaki, Koei; Sakaba, Nariaki; Ohashi, Hirofumi; Yan, X. L.

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. L.

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

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

 Times Cited Count:0 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. L.

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.

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

Uncertainty analysis for source term evaluation of high temperature gas-cooled reactor under accident conditions; Identification of influencing factors in loss-of-forced circulation accidents

Honda, Yuki; Sato, Hiroyuki; Nakagawa, Shigeaki; Ohashi, Hirofumi

Journal of Nuclear Engineering and Radiation Science, 4(3), p.031013_1 - 031013_11, 2018/07

There is growing interest in uncertainty analysis for probabilistic risk assessment (PRA). The focus of this research is to propose and trial investigate the new approach which identify influencing factors for uncertainty in a systematic manner for High Temperature Gas -cooled Reactor (HTGR). As a trial investigation, this approach is tested to evaluation of maximum fuel temperature in a depressurized loss-of-forced circulation (DLOFC) accident and failure of mitigation systems such as control rod systems from the view point of reactor dynamics and thermal hydraulic characteristics. As a result, 16 influencing factors are successfully selected in accordance with the suggested procedure. In the future, the selected influencing factors will be used as input parameter for uncertainty propagation analysis.

JAEA Reports

Comparison between HTFP code and minory changed FORNAX-A code

Aihara, Jun; Ueta, Shohei; Goto, Minoru; Inaba, Yoshitomo; Shibata, Taiju; Ohashi, Hirofumi

JAEA-Technology 2018-002, 70 Pages, 2018/06

JAEA-Technology-2018-002.pdf:1.46MB

HTFP code is code for calculation of additional release amount of fission product (FP) from fuel rod in high temperature gas-cooled reactor (HTGR) after stop of fission. Minory changed Fornax-A code also can calculate that. Therefore, release behavior of Cs calculated with HTFP code was compared with that calculated with minory modified FORNAX-A code in this report. Release constants of Cs evaluated with minory modified FORNAX-A code are rather different from default values for HTFP code.

Journal Articles

Optimization of disposal method and scenario to reduce high level waste volume and repository footprint for HTGR

Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi; Nishihara, Tetsuo; Tsubata, Yasuhiro; Matsumura, Tatsuro

Annals of Nuclear Energy, 116, p.224 - 234, 2018/06

 Times Cited Count:1 Percentile:73.25(Nuclear Science & Technology)

Optimization of disposal method and scenario to reduce volume of High Level Waste (HLW) and the footprint in a geological repository for High Temperature Gas-cooled Reactor (HTGR) has been performed. It was found that HTGR has great advantages to reducing HLW volume and its footprint, which are high burn-up, high thermal efficiency and pin-in-block type fuel, compared with those of LWR and has potential to reduce those more in the previous study. In this study, the scenario is optimized, and the geological repository layout is designed with the horizontal emplacement based on the KBS-3H concept instead of the vertical emplacement based on KBS-3V concept employed in the previous study. As a result, for direct disposal, the repository footprint can be reduced by 20 % by employing the horizontal without change of the scenario. By extending 40 years for cooling time before disposal, the footprint can be reduced by 50 %. For disposal with reprocessing, the number of canister generation can be reduced by 20 % by extending cooling time of 1.5 years between the discharge and reprocessing. The footprint per electricity generation can be reduced by 80 % by extending 40 years before disposal. Moreover, by employing four-group partitioning technology without transmutation, the footprint can be reduced by 90 % with cooling time of 150 years.

JAEA Reports

Evaluation items to attain safety requirements in fuel and core designs for commercial HTGRs

Nakagawa, Shigeaki; Sato, Hiroyuki; Fukaya, Yuji; Tokuhara, Kazumi; Ohashi, Hirofumi

JAEA-Technology 2017-022, 32 Pages, 2017/09

JAEA-Technology-2017-022.pdf:3.59MB

As for the design of commercial HTGRs, the fuel design, core design, reactor coolant system design, secondary helium system design, decay heat removal system design and confinement system design are very important and quite different from those of LWRs. To contribute the establishment of the safety standards for commercial HTGRs, the evaluation items to attain safety requirements in fuel and core designs were studied. In this study, the excellence features of HTGRs based on passive safety or inherent safety were fully reflected. Additionally, concerning the core design, the stability to spatial power oscillation in reactor core of HTGR was studied. The evaluation items as the result of the study are applicable to the safety design of commercial HTGRs in the future.

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