Okita, Shoichiro; Nagaya, Yasunobu; Fukaya, Yuji
Journal of Nuclear Science and Technology, 58(9), p.992 - 998, 2021/09
Fukaya, Yuji; Goto, Minoru; Nishihara, Tetsuo
Nuclear Engineering and Design, 326, p.108 - 113, 2018/01
Burn-up characteristics and criticality of impurity contained into graphite structure for commercial scale prismatic High Temperature Gas-cooled Reactor (HTGR) have been investigated. For HTGR, of which the core is filled graphite structure, the impurity contained into the graphite has unignorable poison effect for criticality. Then, GTHTR300, commercial scale HTGR, employed high grade graphite material named IG-110 to take into account the criticality effect for the reflector blocks next to fuel blocks. The fuel blocks, which should also employ IG-110, employ lower grade graphite material named IG-11 from the economic perspective. In this study, the necessity of high grade graphite material for commercial scale HTGR is reconsidered by evaluating the burn-up characteristics and criticality of the impurity. The poison effect of the impurity, which is used to be expressed by a boron equivalent, reduces exponentially like burn-up of B, and saturate at a level of 1 % of the initial value of boron equivalent. On the other hand, the criticality effect of the boron equivalent of 0.03 ppm, which corresponds to a level of 1 % of IG-11 shows ignorable values lower than 0.01 %k/kk' for both of fuel blocks and reflector blocks. The impurity can be represented by natural boron without problem. Therefore, the poison effect of the impurity is evaluated with whole core burn-up calculations. As a result, it is concluded that the impurity is not problematic from the viewpoint of criticality for commercial scale HTGR because it is burned clearly until End of Cycle (EOC) even with the low grade graphite material of IG-11. According to this result, more economic electricity generation with HTGR is expected by abolishing the utilization of IG-110.
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.
Ota, Masayuki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara
Fusion Engineering and Design, 98-99, p.1847 - 1850, 2015/10
International Reactor Dosimetry and Fusion File release 1.0 (IRDFF 1.0), has been released from the International Atomic Energy Agency (IAEA) recently. In order to validate and test IRDFF 1.0, IAEA has initiated a new Co-ordinated Research Project (CRP). Under this CRP, we have performed an integral experiment on a graphite pseudo-cylindrical slab assembly with DT neutron source at JAEA/FNS. The graphite assembly of 31.4 cm in equivalent radius and 61 cm in thickness is placed at a distance of about 20 cm from the DT neutron source. A lot of foils for the dosimetry reactions in IRDFF1.0 are inserted into the small spaces between the graphite blocks along the center axis of the assembly. After DT neutron irradiation, reaction rates for the dosimetry reactions are measured by the foil activation technique. This experiment is analyzed by using Monte Carlo neutron transport code MCNP5-1.40 with recent nuclear data libraries of ENDF/B-VII.1, JEFF-3.2, and JENDL-4.0. The experimental assembly and DT neutron source are modeled precisely in the MCNP calculation. The reaction rates calculated with IRDFF 1.0 as the response functions for the dosimetry reactions are compared with the experimental values. Also the calculations with JENDL Dosimetry File 99 (JENDL/D-99) are performed for comparison. The results calculated with IRDFF 1.0 show good agreement with the experimental results.
Osaki, Hirotaka; Shimazaki, Yosuke; Sumita, Junya; Shibata, Taiju; Konishi, Takashi; Ishihara, Masahiro
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05
For the design on the VHTR graphite components, it is desirable to employ graphite material with higher strength. IG-430 graphite has been developed as an advanced candidate for VHTR. However, the new developed IG-430 does not have enough databases for the design of HTGR. In this paper, the compressive strength (Cs) of IG-430, one of important strengths for design data, is statistically evaluated. The component reliability is evaluated based on the safety factors defined by the graphite design code, and the applicability as the VHTR graphite material is discussed. It was found that IG-430 has higher strength (about 11%) and lower standard deviation (about 27%) than IG-110 which is one of traditional graphites used for HTGR, because the crack in IG-430 would not easy to propagate rather than IG-110. Since fracture probability for IG-430 is low, the higher reliability of core-component will be achieved using IG-430. It is expected that IG-430 is applicable for VHTR graphite material.
Shimazaki, Yosuke; Isaka, Kazuyoshi; Nomoto, Yasunobu; Seki, Tomokazu; Ohashi, Hirofumi
JAEA-Technology 2014-038, 51 Pages, 2014/12
The analytical models for the evaluation of graphite oxidation were implemented into the THYTAN code, which employs the mass balance and a node-link computational scheme to evaluate tritium behavior in the High Temperature Gas-cooled Reactor (HTGR) systems for hydrogen production, to analyze the graphite oxidation during the air or water ingress accidents in the HTGR systems. This report describes the analytical models of the THYTAN code in terms of the graphite oxidation analysis and its verification and validation (V&V) results. Mass transfer from the gas mixture in the coolant channel to the graphite surface, diffusion in the graphite, graphite oxidation by air or water, chemical reaction and release from the primary circuit to the containment vessel by a safety valve were modeled to calculate the mass balance in the graphite and the gas mixture in the coolant channel. The computed solutions using the THYTAN code for simple questions were compared to the analytical results by a hand calculation to verify the algorithms for each implemented analytical model. A representation of the graphite oxidation experimental was analyzed using the THYTAN code, and the results were compared to the experimental data and the computed solutions using the GRACE code, which was used for the safety analysis of the High Temperature Engineering Test Reactor (HTTR), in regard to corrosion depth of graphite and oxygen concentration at the outlet of the test section to validate the analytical models of the THYTAN code. The comparison of THYTAN code results with the analytical solutions, experimental data and the GRACE code results showed the good agreement.
Ueta, Shohei; Sumita, Junya; Shibata, Taiju; Aihara, Jun; Fujita, Ichiro*; Ohashi, Jun*; Nagaishi, Yoshihide*; Muto, Takenori*; Sawa, Kazuhiro; Sakaba, Nariaki
Nuclear Engineering and Design, 271, p.309 - 313, 2014/05
A new concept of the high temperature gas-cooled reactor (HTGR) is proposed as a challenge to assure no event sequences to the harmful release of radioactive materials even when the design extension conditions (DECs) occur by deterministic approach based on the inherent safety features of the HTGR. The air/water ingress accident, one of the DECs for the HTGR, is prevented by additional measures (e.g. facility for suppression to air ingress). With regard to the core design, it is important to prevent recriticality accidents by keeping the geometry of the fuel rod which consists of the graphite sleeve, fuel compact and SiC-TRISO (TRIstructural-ISOtropic) coated fuel particle, and by improving the oxidation resistance of the graphite when air/water ingress accidents occur. Therefore, it is planned to develop the oxidation-resistant graphite, which is coated with gradient SiC layer. It is also planned that the experimental identification of the condition to form the stable oxide layer (SiO) for SiC layer on the oxidation-resistant graphite and on the SiC-TRISO fuel. This paper describes the R&D plan for un-irradiation and irradiation test under simulating air/water ingress accident condition to develop oxidation-resistant graphite and to investigate the oxidation behavior of SiC coated fuel particle.
Sekiguchi, Tetsuhiro; Baba, Yuji; Shimoyama, Iwao; Nath, K. G.
Surface and Interface Analysis, 38(4), p.352 - 356, 2006/04
We investigated the orientation nature at the top-most layers of F-irradiated graphite using polarization dependent near-edge X-ray absorption fine structure (NEXAFS) spectroscopy which incorporates partial electron yield (PEY) detection and photon-stimulated ion desorption (PSID) techniques. The fluorine K-edge NEXAFS spectra conducted in PEY mode show no significant dependence on polarization angles. In contrast, NEXAFS spectra recorded in F ion yield mode show enhanced yields at a feature of 689.4 eV assigned as a *(C-F) state relevant to =C-F sites, which depend on polarization angles. The C-F bonds prefer relatively tilting down the surface at the top-most layer, while the C-F bonds are randomly directed at deeper regions. We conclude that the difference in the orientation structures between the top surface and bulk is reflected in the NEXAFS recorded in the two different detection modes. It was also found that H- and F- PSID NEXAFS spectra are helpful in understanding desorption mechanism, thus in analysing NEXAFS data.
Sumita, Junya; Shibata, Taiju; Ishihara, Masahiro; Iyoku, Tatsuo; Tsuji, Nobumasa*
Key Engineering Materials, 297-300, p.1698 - 1703, 2005/11
no abstracts in English
JAERI-Tech 2005-048, 108 Pages, 2005/09
The graphite-moderated power reactor was shut down in 1998 and its decommissioning program is being planned. Various graphites are used in the core of magnox-type reactors and HTTR as core-support structural materials and moderating materials of fast neutrons. For the nuclear graphite disposal, it is necessary to determine especially the treatment of long-lived nuclides, such as C which are generated in the graphite components during reactor operation. As a research, which solves the problem of the C concentration, the cooperative research is concluded between JAERI and Japan Nuclear Power Corp. in 1999, and the research for the basic data acquisition has been advanced up to the present. To find the optimum conditions for C reduction, basic data on oxidation reaction and the structure of graphite materials are indispensable. In the present experiment, we measure the air oxidation characteristics in the temperature range 450800C in Quality1 graphite and IG-110 graphite. Changes in pore diameter and pore size distribution due to air oxidation are discussed.
Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Hanawa, Satoshi; Iyoku, Tatsuo; Ishihara, Masahiro
Transactions of 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT-18), p.4822 - 4828, 2005/08
Graphite materials are used for structural components in High Temperature Gas-Cooled Reactor (HTGR) core because of their excellent thermo/mechanical properties. Thermal conductivity of graphite components is reduced by neutron irradiation in reactor operation. The reduced conductivity is expected to be recovered by thermal annealing effect when irradiated graphite component is heated above irradiated temperature. In the present study, temperature analyses considering the annealing effect of the HTGR core at a depressurization accident were carried out and influence of annealing effect on maximum fuel temperature was investigated. The analyses show that the annealing effect can reduce the fuel temperature about 100C at the maximum, and it is possible to evaluate the maximum fuel temperature more appropriately. It was also shown that the core-temperature of High Temperature Engineering Test Reactor (HTTR) at the safety demonstration tests can be analyzed with the developed evaluation method considering annealing effect.
Sawa, Kazuhiro; Ueta, Shohei; Shibata, Taiju; Sumita, Junya; Ohashi, Jumpei; Tochio, Daisuke
JAERI-Tech 2005-024, 34 Pages, 2005/03
The Very-High-Temperature Reactor (VHTR) is one of the strong candidates for the Generation IV Nuclear Energy System. JAERI has developed Zirconium carbide (ZrC)-coated fuel particle and ZrC coating layer is expected to maintain its intactness under higher temperature and burn-up comparing conventional SiC-coating layer. JAERI carries out (1) ZrC-coating process development by large-scale coater, (2) inspection method development and (3) irradiation test and post irradiation experiment of ZrC coated particles. Also, JAERI carries out reactivity insertion tests to clarify the coating failure mechanism and tries to increase allowable temperature limit in case of reactivity insertion accident. Furthermore, JAERI develops non-destructive evaluation methods for mechanical properties of graphite components by ultrasonic testing and micro-indentation technique. This report describes these research and development plan and results of FY 2004 as a MEXT contact research.
Kimura, Hiromi*; Sasaki, Masayoshi*; Morimoto, Yasutomi*; Takeda, Tsuyoshi*; Kodama, Hiroshi*; Yoshikawa, Akira*; Oyaizu, Makoto*; Takahashi, Koji; Sakamoto, Keishi; Imai, Tsuyoshi; et al.
Journal of Nuclear Materials, 337-339, p.614 - 618, 2005/03
no abstracts in English
Hanawa, Satoshi; Ishihara, Masahiro; Motohashi, Yoshinobu*
Zairyo, 54(2), p.201 - 206, 2005/02
no abstracts in English
Goniche, M.*; Kazarian, F.*; Bibet, P.*; Maebara, Sunao; Seki, Masami; Ikeda, Yoshitaka; Imai, Tsuyoshi*
Journal of Vacuum Science and Technology A, 23(1), p.55 - 65, 2005/01
Outgassing rates have been measured for long duration (100-4700 seconds) of RF transmission at high power density (50-200 MW/m) for waveguides made of OFHC copper, dispersoid copper, copper-coated carbon fiber composite and copper-coated graphite. The measurements were performed on multi-waveguide(2 to 8)mockups, using a test bed facility equipped with a 3.7 GHz klystron. The effect on the outgassing rate of waveguide surface temperature and of initial wall gas loading('conditioning'), is examined. It is concluded that an outgassing rate of 110 Pamsm and 510 Pamsm at 300C and 400C respectively, can be expected for the tested material. Based on these measurement results, it is further concluded that no additional pumping will be needed for the LHRF antenna proposed for ITER.
Sumita, Junya; Ishihara, Masahiro; Nakagawa, Shigeaki; Kikuchi, Takayuki; Iyoku, Tatsuo
Nuclear Engineering and Design, 233(1-3), p.81 - 88, 2004/10
A High Temperature Gas-cooled Reactor is particularly attractive due to its capability of producing high temperature helium gas and its possibility to exploit inherent safety characteristic. To achieve high temperature helium-gas, reactor internals are made of graphite and heat resistant materials, its surroundings are composed of metals. The reactor internals of the HTTR consist of graphite and metallic core support structures and shielding blocks. This paper describes the reactor internal design of the HTTR, especially the core support graphite structures, and the program of an in-service inspection.
Ishihara, Masahiro; Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Oku, Tatsuo*
Nuclear Engineering and Design, 233(1-3), p.251 - 260, 2004/10
The High Temperature Engineering Test Reactor (HTTR) constructed by Japan Atomic Energy Research Institute (JAERI) is a graphite-moderated and helium-gas-cooled reactor with prismatic fuel elements of hexagonal blocks. The reactor internal structures of the HTTR are mainly made up of graphite components. As well known, the graphite is a brittle material and there were no available design criteria for brittle materials. Therefore, JAERI had to develop the design criteria taking account of the brittle fracture behavior. In this paper, concept and key specification of the developed graphite design criteria is described, and also an outline of the quality control specified in the design criteria is mentioned.
Sumita, Junya; Nakano, Masaaki*; Tsuji, Nobumasa*; Shibata, Taiju; Ishihara, Masahiro
JAERI-Tech 2004-055, 25 Pages, 2004/08
Neutron irradiation remarkably reduces the thermal conductivity of graphite, and the reduced thermal conductivity is recovered by annealing effect if the graphite is heated above the irradiation temperature. Therefore, it is expected that the reduced thermal conductivity of graphite components in the HTGR could be recovered by the annealing effect in accidents, such as a depressurization accident. Then, an analytical investigation of the annealing effect on thermal performance of a HTGR core was carried. The analysis showed that the annealing effect reduces the maximum fuel temperature about 70C, and it is important to introduce the annealing effect appropriately in the temperature analysis of the core components and reactor internals. In addition, an annealing test method was investigated to evaluate the effect quantitatively, and the test plan was made.
Nath, K. G.; Shimoyama, Iwao; Sekiguchi, Tetsuhiro; Baba, Yuji
Applied Surface Science, 234(1-4), p.234 - 239, 2004/07
Here we report oxidization properties of Si nanostructures grown on graphite. Si 1s X-ray photoemission spectra using synchrotron radiation are used in order to understand the oxidization pathways. Several Si films, such as 0.4, 2, 5.5 & Aring; were grown on highly oriented pyrolitic graphite (HOPG). In the case of a 0.4 & Aring; Si on HOPG, where different types of Si nanostructures in the form of nanoclusters are present, oxygen reactivity is nearly zero. In contrast, the thick film (5.5 & Aring;), where a bulk-type phase is present, shows a higher degree of reactivity. The results are discussed on the basis of nanostructure geometry, number of constituting Si atoms and cluster size.
Sumita, Junya; Shibata, Taiju; Baba, Shinichi; Ishihara, Masahiro; Iyoku, Tatsuo
Nihon Kikai Gakkai M&M 2004 Zairyo Rikigaku Kanfarensu Koen Rombunshu, p.141 - 142, 2004/07
Neutron irradiation reduces the thermal conductivity of graphite, but the reduced thermal conductivity is recovered by annealing effect if the graphite is heated above the irradiation temperature. In this research, annealing effect of thermal conductivity on thermal stress of graphite is investigated. In addition, effect of recovered thermal conductivity on membrane, point and peak stress is also investigated.