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Aihara, Jun; Goto, Minoru; Inaba, Yoshitomo; Ueta, Shohei; Sumita, Junya; Tachibana, Yukio
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.814 - 822, 2016/11
Japan Atomic Energy Agency (JAEA) has started R&D for apply SiC/C mixed matrix to fuel element of high temperature gas-cooled reactors (HTGRs) to improve oxidation resistance of fuel. Nuclear thermal design of HTGR with SiC/C mixed matrix fuel compacts was carried out as a part of above R&Ds. Nuclear thermal design was carried out based on a small sized HTGR for developing countries, HTR50S. Maximum enrichment of uranium is set to be 10 wt%, because coated fuel particles with 10 wt% uranium have been fabricated in Japan. Numbers of kinds of enrichment and burnable poisons (BPs) were set to be same as those of original HTR50S (3 and 2, respectively). We succeeded in nuclear thermal design of a small sized HTGR which performance was equivalent to original HTR50S, with SiC/C mixed matrix fuel compacts. Based on nuclear thermal design, intactness of coated fuel particles was evaluated to be kept on internal pressure during normal operation.
test plant; System performance evaluation for HTTR gas turbine cogeneration plantSato, Hiroyuki; Nomoto, Yasunobu; Horii, Shoichi; Sumita, Junya; Yan, X.; Ohashi, Hirofumi
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.759 - 766, 2016/11
This paper presents the system performance evaluation for HTTR gas turbine cogeneration test plant (HTTR-GT/H 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 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 commercial Gas Turbine High Temperature Reactor Cogeneration (GTHTR300C). The results also lead us to the conclusion that HTTR-GT/H plant can be used to test operational procedure unique to HTGR direct-cycle gas turbine cogeneration.
Honda, Yuki; Fukaya, Yuji; Nakagawa, Shigeaki; Baker, R. I.*; Sato, Hiroyuki
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.704 - 713, 2016/11
A high-temperature gas-cooled reactor (HTGR) has superior safety characteristics. A loss of forced cooling (LOFC) test using a high-temperature engineering test reactor (HTTR) has been carried out to verify the inherent safety of an HTGR when forced cooling is diminished without reactor scram. In the test, an all-gas circulator was tripped with an initial reactor power of 9 MW and re-criticality was shown. This study focuses on developing a point kinetics method with RELAP5-3D code for an LOFC accident. There is a large temperature difference between the inlet and outlet of the core in an HTGR, and the temperature fluctuation range has been large in several accidents. We analyze the temperature dependency of xenon-135 reactivity and show that the temperature dependency of xenon-135 microscopic absorption cross-section affected the re-criticality time of the LOFC test.
SO
decomposition, HI distillation, and HI decomposition sectionNoguchi, Hiroki; Takegami, Hiroaki; Kamiji, Yu; Tanaka, Nobuyuki; Iwatsuki, Jin; Kasahara, Seiji; Kubo, Shinji
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.1029 - 1038, 2016/11
JAEA has been conducting R&D on the IS process for nuclear-powered hydrogen production. We have constructed a 100 NL/h-H
-scale test apparatus made of industrial materials. At first, we investigated performance of components in this apparatus. In this paper, the test results of HSO
decomposition, HI distillation, and HI decomposition were shown. In the HSO section, O production rate is proportional to HSO feed rate and SO
decomposition ratio was estimated about 80%. In HI distillation section, we confirmed to acquire a concentrated HI solution over azeotropic HI composition in the condenser. In HI decomposition section, H could be produced stably by HI decomposer and decomposition ratio was about 18%. The HSO decomposer, the HI distillation column, and the HI decomposer were workable. Based on the results added to that shown in Series I, we conducted a trial continuous operation and succeeded it for 8 hours.
Tanaka, Nobuyuki; Takegami, Hiroaki; Noguchi, Hiroki; Kamiji, Yu; Iwatsuki, Jin; Aita, Hideki; Kasahara, Seiji; Kubo, Shinji
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.1022 - 1028, 2016/11
Japan Atomic Energy Agency (JAEA) has manufactured 100 NL/h-H-scale hydrogen test apparatus. In advance to conduct the continuous operation, we investigated performance of the components in each section of the IS process. In this paper, the results of test of Bunsen and HI concentration sections was shown. In Bunsen reaction, section, we confirmed that outlet gas flow rate included no SO gas, indicating that all the feed SO gas was absorbed to the solution in the Bunsen reactor for the Bunsen reaction. On the basis of these results, we evaluated that Bunsen reactor was workable. In HI concentration section, HI concentration was conducted by EED stack. As a result, it can concentrate HI in HIx solution as theoretically predicted on the basis of the previous paper. Based on the results added to that shown in Series II, we have conducted a trial continuous operation and succeeded it for 8 hours.
Ohashi, Hirofumi; Sato, Hiroyuki; Nakagawa, Shigeaki; Tokuhara, Kazumi; Nishihara, Tetsuo; Kunitomi, Kazuhiko
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.330 - 340, 2016/11
The safety requirements for the design of HTGRs has been developed by the research committee established in the Atomic Energy Society of Japan so as to incorporate the HTGR safety features demonstrated by HTTR, lessons learned from the accident of Fukushima Daiichi Nuclear Power Station and requirements for the coupling of the hydrogen production plants with nuclear plant. The safety design approach was determined to establish a high level of safety design standards by utilizing inherent safety features of HTGRs. This paper describes the process to develop the HTGR specific safety requirements and overview of the proposed HTGR specific safety requirements.
Reitsma, F.*; Kunitomi, Kazuhiko; Ohashi, Hirofumi
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.341 - 352, 2016/11
The IAEA Coordinated Research Project (CRP) on Modular High Temperature Gas-cooled Reactor (HTGR) Safety Design started in December 2014 and is now in its second year of implementation. The project investigates and makes proposals on safety design criteria by not only considering the existing water cooled reactors safety requirements for nuclear power plant design but also making use of past and current experience of HTGR licensing in the IAEA member states including Japan. The detailed project content and a description of modular HTGRs and its safety features are given. This informs the comments made on the applicability of the existing requirements and the examples of newly proposed HTGR specific safety requirements under development.
Inaba, Yoshitomo; Honda, Yuki; Nishihara, Tetsuo
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.985 - 990, 2016/11
In order to ensure the thermal integrity of fuel in high temperature gas-cooled reactors (HTGRs), it is necessary that the maximum fuel temperature in the normal operation is to be lower than the thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as thermal power, core geometry, power and neutron fluence distributions, and core coolant flow distribution. The core coolant flow distribution calculation code used in the design stage of High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and not user-friendly. Therefore, a new core coolant flow distribution calculation code with a user-friendly system such as simple and easy operations and execution procedures has been developed. In this paper, the outline of the new code is described and the simulation result of an out-of-pile test with one fuel column is shown as the first step of the code validation. The simulation results provide good agreement with the test one.
Shibata, Taiju; Sumita, Junya; Sakaba, Nariaki; Osaki, Takashi*; Kato, Hideki*; Izawa, Shoichi*; Muto, Takenori*; Gizatulin, S.*; Shaimerdenov, A.*; Dyussambayev, D.*; et al.
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.567 - 571, 2016/11
Graphite are used for the in-core components of HTGR, and it is desirable to enhance oxidation resistance to keep much safety margin. SiC coating is the candidate method for this purpose. JAEA and four Japanese graphite companies are studying to develop oxidation-resistant graphite. Neutron irradiation test was carried out by WWR-K reactor of INP of Kazakhstan through ISTC partner project. The total irradiation cycles of WWR-K operation was 10 cycles by 200 days. Irradiation temperature about 1473 K would be attained. The maximum fast neutron fluence (E 
0.18 MeV) for the capsule irradiated at a central irradiation hole was preliminary calculated as 1.2
10
/m
, and for the capsule at a peripheral irradiation hole as 4.210
/m. Dimension and weight of the irradiated specimens were measured, and outer surface of the specimens were observed by optical microscope. For the irradiated oxidation resistant graphite, out-of-pile oxidation test will be carried out at an experimental laboratory.
Chikhray, Y.*; Kulsartov, T.*; Shestakov, V.*; Kenzhina, I.*; Askerbekov, S.*; Sumita, Junya; Ueta, Shohei; Shibata, Taiju; Sakaba, Nariaki; Abdullin, Kh.*; et al.
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.572 - 577, 2016/11
Application of SiC as corrosion-resistive coating over graphite remains important task for HTGR. This study presents the results of chemical interaction of the SiC gradient coating over the high-density IG-110 graphite with water vapor in the temperature up to 1673 K. The experiments at 100 Pa of water vapor showed that the passive reaction caused to form SiO film on the surface of SiC coating. Active corrosion of SiC in 1Pa of water vapor leads to deposits of various carbon composites on its surface.
Ueta, Shohei; Aihara, Jun; Shaimerdenov, A.*; Dyussambayev, D.*; Gizatulin, S.*; Chakrov, P.*; Sakaba, Nariaki
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.246 - 252, 2016/11
In order to examine irradiation performance of the new Tri-structural Isotropic (TRISO) fuel for the High Temperature Gas-cooled Reactor (HTGR) at the burnup around 100 GWd/t, a capsule irradiation test was conducted by WWR-K research reactor in the Institute of Nuclear Physics (INP) of Kazakhstan. The irradiated TRISO fuel was designed by Japan Atomic Energy Agency (JAEA) and fabricated in basis of the HTTR fuel technology in Japan. The fractional release of fission gas from the fuel during the irradiation shows good agreement with the predicted one released from as-fabricated failed TRISO fuel. It was suggested that unexpected additional fuel failure would not occur during the irradiation up to 100 GWd/t. In addition, the post-irradiation examination (PIE) with the irradiated fuel is planned to qualify TRISO fuel integrity and upgrade HTGR fuel design for further burnup extension.
Kasahara, Seiji; Imai, Yoshiyuki; Suzuki, Koichi*; Iwatsuki, Jin; Terada, Atsuhiko; Yan, X.
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.491 - 500, 2016/11
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 Japan Atomic Energy Agency's commercial Gas Turbine High Temperature Reactor Cogeneration (GTHTR300C) plant design. Innovative techniques proposed by JAEA were applied for improvement of hydrogen production thermal efficiency; flash concentration of HSO using waste heat from Bunsen reaction, prevention of HSO vaporization from a distillation column by introduction of HSO solution, and I condensation heat recovery by direct contact heat exchange in the HI distillation column. A simulation of material and heat balance showed hydrogen of about 31,900 Nm
/h was produced by 170 MW heat from the GTHTR300C. A process thermal efficiency of 50.2% was achievable with incorporation of the innovative techniques and several high performance components expected in future R&D.
test plant; System designYan, X.; Sato, Hiroyuki; Sumita, Junya; Nomoto, Yasunobu; Horii, Shoichi; Imai, Yoshiyuki; Kasahara, Seiji; Suzuki, Koichi*; Iwatsuki, Jin; Terada, Atsuhiko; et al.
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.827 - 836, 2016/11
Pre-licensing basic design for a cogenerating HTGR test plant system is presented. The plant to be coupled to existing 30 MWt 950
C test reactor HTTR is intended as a system technology demonstrator for GTHTR300C plant design. More specifically the test plant of HTTR-GT/H aims to (1)demonstrate the licensability of the GTHTR300C for electricity production by gas turbine and hydrogen cogeneration by thermochemical process and (2) confirm the operation control and safety of such cogeneration system. With construction and operation completion by 2025, the test plant is expected to be the first of a kind HTGR-powered cogeneration plant operating on the two advanced energy conversion systems of closed cycle helium gas turbine for power generation and thermochemical iodine-sulfur water-splitting process for hydrogen production.
Kunitomi, Kazuhiko
no journal, ,
This presentation describes the status of HTTR project including development of hydrogen production technology, etc., re-start of HTTR. Also, governmental activities on the HTGR development in Japan are shown.