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Shaimerdenov, A.*; Gizatulin, S.*; Dyussambayev, D.*; Askerbekov, S.*; Ueta, Shohei; Aihara, Jun; Shibata, Taiju; Sakaba, Nariaki
Nuclear Engineering and Technology, 54(8), p.2792 - 2800, 2022/08
Times Cited Count：7 Percentile：88.9(Nuclear Science & Technology)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.
Aihara, Jun; Ueta, Shohei; Honda, Masaki*; Blynskiy, P.*; Gizatulin, S.*; Sakaba, Nariaki; Tachibana, Yukio
Journal of Nuclear Science and Technology, 51(11-12), p.1355 - 1363, 2014/11
Times Cited Count：10 Percentile：60.98(Nuclear Science & Technology)Plan and status of research and development (R&D) were described on coated fuel particle (CFP) and fuel compacts for core of small sized high temperature gas-cooled reactor (HTGR) HTR50S at 2nd step of phase I (second core of HTR50S). Specifications of existing CFPs for high burnup (HTR50S2-type-CFPs) were adopted as specifications of CFPs, to reduce the R&D. HTR50S2-type-CFPs were fabricated based on technology developed in High Temperature Engineering Test Reactor (HTTR) project. First irradiation test of HTR50S2-type-CFPs is now being carried out. In addition, R&D for fuel compact with high packing fraction is needed, because volume fraction of fuel kernel to whole of HTR50S2-type-CFP is rather smaller than that of the HTTR-type-CFP. In addition, we describe outline of R&D plans for core of HTR50S in phase II and naturally safe HTGR.
Ueta, Shohei; Shaimerdenov, A.*; Gizatulin, S.*; Chekushina, L.*; Honda, Masaki*; Takahashi, Masashi*; Kitagawa, Kenichi*; Chakrov, P.*; Sakaba, Nariaki
Proceedings of 7th International Topical Meeting on High Temperature Reactor Technology (HTR 2014) (USB Flash Drive), 7 Pages, 2014/10
A capsule irradiation test with the high temperature gas-cooled reactor (HTGR) fuel is being carried out using WWR-K research reactor in the Institute of Nuclear Physics of the Republic of Kazakhstan (INP) to attain 100 GWd/t-U of burnup under normal operating condition of a practical small-sized HTGR. This is the first HTGR fuel irradiation test for INP in Kazakhstan collaborated with Japan Atomic Energy Agency (JAEA) in frame of International Science and Technology Center (ISTC) project. In the test, TRISO coated fuel particle with low-enriched UO
(less than 10% of 
U) is used, which was newly designed by JAEA to extend burnup up to 100 GWd/t-U comparing with that of the HTTR (33 GWd/t-U). Both TRISO and fuel compact as the irradiation test specimen were fabricated in basis of the HTTR fuel technology by Nuclear Fuel Industries, Ltd. in Japan. A helium-gas-swept capsule and a swept-gas sampling device installed in WWR-K were designed and constructed by INP. The irradiation test has been started in October 2012 and will be completed up to the end of February 2015. The irradiation test is in the progress up to 69 GWd/t of burnup, and integrity of new TRISO fuel has been confirmed. In addition, as predicted by the fuel design, fission gas release was observed due to additional failure of as-fabricated SiC-defective fuel.
Mo manufacturing process, 1; Reusability evaluation of Mo absorbent (Joint research)Kimura, Akihiro; Niizeki, Tomotake*; Kakei, Sadanori*; Chakrova, Y.*; Nishikata, Kaori; Hasegawa, Yoshio*; Yoshinaga, Hideo*; Chakrov, P.*; Tsuchiya, Kunihiko
JAEA-Technology 2013-025, 40 Pages, 2013/10
JAEA-Technology-2013-025.pdf:2.62MB
Neutron Irradiation and Testing Reactor Center has developed the production of a medical isotope of Mo, the parent nuclide of 
Tc by the (n,
) method using JMTR. The (n,) method has an advantage of easy manufacturing process and low radioactive wastes generation. However, the low radioactivity concentration of Tc is remaining as an issue. Therefore, PZC and PTC have been developed as adsorbent of molybdenum. Meanwhile, it is necessary to recycle the absorbent and Mo for the reduction of the radioactive waste of used-adsorbent and the effective use of limited resources, respectively. This report summarizes results of the synthesis of Mo adsorbents such as PZC and PTC, and the performance tests.
Dorn, C. K.*; Tsuchiya, Kunihiko; Takemoto, Noriyuki; Ito, Masayasu; Hori, Junichi*; Chekushina, L.*; Hatano, Yuji*; Chakrov, P.*; Kawamura, Hiroshi
Proceedings of 6th International Symposium on Material Testing Reactors (ISMTR-6) (Internet), 9 Pages, 2013/10
no abstracts in English
Mo-Tc production development by (n,) reactionTsuchiya, Kunihiko; Mutalib, A.*; Chakrov, P.*; Kaminaga, Masanori; Ishihara, Masahiro; Kawamura, Hiroshi
JAEA-Conf 2011-003, p.137 - 141, 2012/03
As one of effective uses of the JMTR, JAEA has a plan to produce Mo by (n,) method, a parent nuclide of Tc. In case of Japan, the supplying of Mo depends only on imports from foreign countries, the R&D on production method of Mo-Tc has been performed with foreign countries and Japanese industrial users under the cooperation programs. The main R&D items for the production are (1) Fabrication of irradiation target such as the sintered MoO
pellets, (2) Separation and concentration of Tc by the solvent extraction from Mo solution, (3) Examination of Tc solution for a medicine, and (4) Mo recycling from Mo generator and solution. Especially, it is important to establish the separation and extraction methods in the item (2) and the experiments and information exchanges in some methods have been carried out under the international cooperation. In this paper, the status of the R&D is introduced for the production of Mo-Tc.
Dorn, C. K.*; Tsuchiya, Kunihiko; Hatano, Yuji*; Chakrov, P.*; Kodama, Mitsuo*; Kawamura, Hiroshi
JAEA-Conf 2011-003, p.93 - 97, 2012/03
The JMTR has used beryllium reflector since it began operation in 1968. Beryllium has been used as the reflector element material in the reactor, specifically S-200F structural grade beryllium in JMTR. As a part of the reactor upgrade, the Japan Atomic Energy Agency (JAEA) has carried out the cooperation experiments to extend the operating lifetime of the beryllium reflector elements. Thus, three kinds of beryllium metals such as S-200F, S-65H and I-220H were selected at the viewpoints of production methods, impurities and grain size of beryllium starting powders, mechanical properties. Now, data of the material properties of these beryllium grades are accumulated under un-irradiated and irradiated conditions. In this paper, the results of various properties and irradiation test plan for lifetime expansion of beryllium are described for material testing reactors.
Kips, R.*; Lindvall, R.*; Marks, N.*; Gluschenko, V.*; Okubo, Ayako; Szeles, E.*
no journal, ,
