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Journal Articles

Study on the effect of long-term high temperature irradiation on TRISO fuel

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:4 Percentile:90.16(Nuclear Science & Technology)

Journal Articles

Post irradiation experiment about SiC-coated oxidation-resistant graphite for high temperature gas-cooled reactor

Shibata, Taiju; Mizuta, Naoki; Sumita, Junya; Sakaba, Nariaki; Osaki, Takashi*; Kato, Hideki*; Izawa, Shoichi*; Muto, Takenori*; Gizatulin, S.*; Shaimerdenov, A.*; et al.

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

Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR). Oxidation damage on the graphite components in air ingress accident is a crucial issue for the safety point of view. SiC coating on graphite surface is a possible technique to enhance oxidation resistance. However, it is important to confirm the integrity of this material against high temperature and neutron irradiation for the application of the in-core components. JAEA and Japanese graphite companies carried out the R&D to develop the oxidation-resistant graphite. JAEA and INP investigated the irradiation effects on the oxidation-resistant graphite by using a framework of ISTC partner project. This paper describes the results of post irradiation experiment about the neutron irradiated SiC-coated oxidation-resistant graphite. A brand of oxidation-resistant graphite shows excellent performance against oxidation test after the irradiation.

Journal Articles

Investigation of irradiated properties of extended burnup TRISO fuel

Shaimerdenov, A.*; Gizatulin, S.*; Kenzhin, Y.*; Dyussambayev, D.*; Ueta, Shohei; Aihara, Jun; Shibata, Taiju

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

The Institute of Nuclear Physics of the Republic of Kazakhstan (INP) conducts an irradiation test and post-irradiation examinations (PIEs) of the high-temperature gas-cooled reactor (HTGR) fuel and materials to develop the extend burnup fuel up to 100 GWd/t-U collaboratively with the Japan Atomic Energy Agency (JAEA) under projects in a frame of the International Science and Technology Centre (ISTC). Cylindrical fuel compact specimens consisting of newly-designed TRISO (tri-structural isotropic) coated fuel particles and a matrix made of graphite material were manufactured in Japan. An irradiation test of the fuel specimens using a helium-gas swept capsule designed and constructed in the INP has been performed up to 100 GWd/t-U in the WWR-K research reactor by April 2015. In the next stage, PIEs with the irradiated fuel specimens have been started in February 2017 as a new ISTC project. Several PIE technologies by non-destructive and destructive techniques with irradiated fuel compacts were developed by the INP. This report presents the developed technologies and interim results of the PIE for high burning TRISO fuel.

Journal Articles

Enhancement of oxidation tolerance of graphite materials for high temperature gas-cooled reactor

Mizuta, Naoki; Sumita, Junya; Shibata, Taiju; Osaki, Takashi*; Kato, Hideki*; Izawa, Shoichi*; Muto, Takenori*; Gizatulin, S.*; Sakaba, Nariaki

Tanso Zairyo Kagaku No Shinten; Nihon Gakutsu Shinkokai Dai-117-Iinkai 70-Shunen Kinen-Shi, p.161 - 166, 2018/10

To enhance oxidation resistance of graphite material for in-core components of HTGR, JAEA and four Japanese graphite companies; Toyo Tanso, IBIDEN, Tokai Carbon and Nippon Techno-Carbon, are carrying out for development of oxidation-resistant graphite by CVD-SiC coating. This paper describes the outline of neutron irradiation test about the oxidation-resistant graphite by WWR-K reactor of INP, Kazakhstan through an ISTC partner project. Prior to the irradiation test, the oxidation-resistant graphite by CVD-SiC coating of all specimens showed enough oxidation resistance under un-irradiation condition. The neutron irradiation test was already completed and out-of-pile oxidation test will be carried out at the hot-laboratory of WWR-K.

Journal Articles

Irradiation test about oxidation-resistant graphite in WWR-K research reactor

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$$times$$10$$^{25}$$/m$$^{-2}$$, and for the capsule at a peripheral irradiation hole as 4.2$$times$$10$$^{24}$$/m$$^{-2}$$. 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.

Journal Articles

Irradiation test and post irradiation examination of the high burnup HTGR fuel

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.

JAEA Reports

Irradiation test with silicon ingot for NTD-Si irradiation technology

Takemoto, Noriyuki; Romanova, N.*; Kimura, Nobuaki; Gizatulin, S.*; Saito, Takashi; Martyushov, A.*; Nakipov, D.*; Tsuchiya, Kunihiko; Chakrov, P.*

JAEA-Technology 2015-021, 32 Pages, 2015/08

JAEA-Technology-2015-021.pdf:3.15MB

Silicon semiconductor production by neutron transmutation doping (NTD) method using the JMTR has been investigated in Neutron Irradiation and Testing Reactor Center, Japan Atomic Energy Agency in order to expand the industry use. As a part of investigations, irradiation test with a silicon ingot was planned using WWR-K in Institute of Nuclear Physics, Republic of Kazakhstan. A device rotating the ingot made with the silicon was fabricated and was installed in the WWR-K for the irradiation test. And that, a preliminary irradiation test was carried out using neutron fluence monitors to evaluate the neutronic irradiation field. Based on the result, two silicon ingots were irradiated as scheduled, and the resistivity of each irradiated silicon ingot was measured to confirm the applicability of high-quality silicon semiconductor by the NTD method (NTD-Si) to its commercial production.

Journal Articles

Development plan of high burnup fuel for high temperature gas-cooled reactors in future

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:61.24(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.

Journal Articles

Irradiation performance of HTGR fuel in WWR-K research reactor

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$$_{2}$$ (less than 10% of $$^{235}$$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.

Journal Articles

Irradiation test plan of oxidation-resistant graphite in WWR-K research reactor

Sumita, Junya; Shibata, Taiju; Sakaba, Nariaki; Osaki, Hiroki*; Kato, Hideki*; Fujitsuka, Kunihiro*; Muto, Takenori*; Gizatulin, S.*; Shaimerdenov, A.*; Dyussambayev, D.*; et al.

Proceedings of 7th International Topical Meeting on High Temperature Reactor Technology (HTR 2014) (USB Flash Drive), 7 Pages, 2014/10

Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor(HTGR)which is a graphite-moderated and helium gas-cooled reactor. In the case of air ingress accident in HTGR, SiO$$_{2}$$ protective layer is formed on the surface of SiC layer in TRISO CFP and oxidation of SiC does not proceed and fission products are retained inside the fuel particle. A new safety concept for the HTGR, called Naturally Safe HTGR, has been recently proposed. To enhance the safety of Naturally Safe HTGR ultimately, it is expected that oxidation-resistant graphite is used for graphite components to prevent the TRISO CFPs and fuel compacts from failure. SiC coating is one of candidate methods for oxidation-resistant graphite. JAEA and four graphite companies launched R&Ds to develop the oxidation-resistant graphite and the International Science and Technology Center(ISTC) partner project with JAEA and INP was launched to investigate the irradiation effects on the oxidation-resistant graphite. To determine grades of the oxidation-resistant graphite which will be adopted as irradiation test, a preliminary oxidation test was carried out. This paper described the results of the preliminary oxidation test, the plan of out-of-pile test, irradiation test and post-irradiation test(PIE)of the oxidation-resistant graphite.

JAEA Reports

Design, fabrication and transportation of Si rotating device

Kimura, Nobuaki; Imaizumi, Tomomi; Takemoto, Noriyuki; Tanimoto, Masataka; Saito, Takashi; Hori, Naohiko; Tsuchiya, Kunihiko; Romanova, N. K.*; Gizatulin, S.*; Martyushov, A.*; et al.

JAEA-Technology 2012-012, 34 Pages, 2012/06

JAEA-Technology-2012-012.pdf:12.91MB

Si semiconductor production by Neutron Transmutation Doping (NTD) method using the Japan Materials Testing Reactor (JMTR) has been investigated in Neutron Irradiation and Testing Reactor Center, Japan Atomic Energy Agency (JAEA) in order to expand industry use. As a part of investigations, irradiation test of silicon ingot for development of NTD-Si with high quality was planned using WWR-K in Institute of Nuclear Physics (INP), National Nuclear Center of Republic of Kazakhstan (NNC-RK) based on one of specific topics of cooperation (STC), Irradiation Technology for NTD-Si (STC No.II-4), on the implementing arrangement between NNC-RK and the JAEA for "Nuclear Technology on Testing/Research Reactors" in cooperation in research and development in nuclear energy and technology. As for the irradiation test, Si rotating device was fabricated in JAEA, and the fabricated device was transported with irradiation specimens from JAEA to INP-NNC-RK. This report described the design, the fabrication, the performance test of the Si rotating device and transportation procedures.

JAEA Reports

Status of international cooperation in nuclear technology on testing/research reactors between JAEA and INP-NNC

Kawamura, Hiroshi; Chakrov, P.*; Tsuchiya, Kunihiko; Gizatulin, S.*; Takemoto, Noriyuki; Chakrova, Y.*; Kimura, Akihiro; Ludmila, C.*; Tanimoto, Masataka; Asset, S.*; et al.

JAEA-Review 2011-042, 46 Pages, 2012/02

JAEA-Review-2011-042.pdf:2.69MB

Based on the implementing agreement between National Nuclear Center of the Republic of Kazakhstan (NNC) and the Japan Atomic Energy Agency (JAEA) for the Nuclear Technology on Testing/Research Reactors in the cooperation in Nuclear Energy Research and Development in Nuclear Energy and Technology, four specific topics of cooperation (STC) have been carried out from June, 2009. Four STCs are as follows; (1) STC No.II-1: International Standard of Instrumentation, (2) STC No.II-2: Irradiation Technology of RI Production, (3) STC No.II-3: Lifetime Expansion of Beryllium Reflector, (4) STC No.II-4: Irradiation Technology for NTD-Si. The information exchange, personal exchange and cooperation experiments are carried out under these STCs. The status in the field of nuclear technology on testing/research reactors in the implementing arrangement is summarized, and future plans of these specific topics of cooperation are described in this report.

Journal Articles

Tritium generation in lithium ceramics Li$$_{2}$$TiO$$_{3}$$ for fusion reactor blanket

Tazhibayeva, I. L.*; Kenzhin, E. A.*; Kulsartov, T. V.*; Kuykabayeva, A. A.*; Shestakov, V.*; Chikhray, E.*; Gizatulin, S.*; Maksimkin, O. P.*; Beckman, I. N.*; Kawamura, Hiroshi; et al.

Questions of Atomic Science and Technology, 2, p.3 - 11, 2008/00

Lithium titanate (Li$$_{2}$$TiO$$_{3}$$) was chosen as a tentative reference material from viewpoints of good tritium recovery at low temperatures and of low tritium inventory and chemical stability for the breeding blanket in fusion reactors. The results of the irradiation tests of Li$$_{2}$$TiO$$_{3}$$ in the WWR-K of NNC-RK are described in this paper. 96at% $$^{6}$$Li-enriched Li$$_{2}$$TiO$$_{3}$$ pebbles and disks were prepared as the irradiation specimens and these specimens were irradiated during 220 days (5350 hours) at the reactor power of 6 MWt. Tritium release was measured continuously during irradiation tests and tritium release properties were evaluated. The mechanics describing generation and release of tritium from the irradiated Li$$_{2}$$TiO$$_{3}$$ were analyzed. There was estimated tritium loss due to recoil energy and binding of tritium in HTO, and there was calculated stationary tritium release due to diffusion under constant temperature and under thermal cycling.

Oral presentation

Collaboration of JAEA and NNC for Kazakhstan project on high-temperature gas-cooled reactor

Nakatsuka, Toru; Levin, A. G.*; Ueta, Shohei; Gizatulin, S.*; Tachibana, Yukio; Kolodeshnikov, A.*; Sakaba, Nariaki; Chakrov, P.*; Kunitomi, Kazuhiko; Vassiliev, Y. S.*; et al.

no journal, , 

The small-sized high-temperature gas-cooled reactors (HTGRs) with an electric power rating of less than 300 MWe can greatly facilitate decentralized energy supply, and create new industries and stimulate economical development in cities and localities as well as in those remote regions to which power transmission grids are undeveloped in developing countries such as Kazakhstan. In 2007, Japan Atomic Energy Agency (JAEA) and National Nuclear Center of Kazakhstan (NNC) have started to collaborate in nuclear energy research and development for early realization of deployment of the HTGR in Kazakhstan, and to support for the Kazakhstan HTGR (KHTR) Project by utilizing the technologies developed under the High Temperature Engineering Test Reactor (HTTR) Project. In 2010, JAEA started a conceptual design of KHTR steam turbine system with thermal power of 50 MW and the maximum coolant temperature at reactor outlet of 750 $$^{circ}$$C for earlier development of HTGRs with support of Japan parties, which consists of Japanese industrial companies, etc. in order to support NNC for preparation of the feasibility study of KHTR.

Oral presentation

Collaboration with Republic of Kazakhstan regarding development of HTGR, 3; Collaboration of development of oxidation-resistant graphite material for HTGR

Shibata, Taiju; Sumita, Junya; Nagata, Hiroshi; Saito, Takashi; Tsuchiya, Kunihiko; Sakaba, Nariaki; Osaki, Hiroki*; Kato, Hideki*; Fujitsuka, Kunihiro*; Muto, Takenori*; et al.

no journal, , 

no abstracts in English

Oral presentation

Collaboration with Republic of Kazakhstan regarding development of HTGR, 2; Collaboration of irradiation performance of HTGR fuel

Ueta, Shohei; Mizutani, Yoshitaka; Sakaba, Nariaki; Furihata, Noboru*; Honda, Masaki*; Asset, S.*; Gizatulin, S.*; Chakrov, P.*

no journal, , 

A capsule irradiation test with the high temperature gas-cooled reactor (HTGR) fuel by WWR-K in the Institute of Nuclear Physics of the Republic of Kazakhstan (INP) is being carried out. The HTGR fuel specimens were newly designed at the target burnup of 100 GWd/t. A plan of the irradiation test and results on evaluation of the integrity of the HTGR fuel specimen based on fission gas (FP) release rate under the irradiation are reported.

Oral presentation

Collaboration with Republic of Kazakhstan regarding irradiation performance of HTGR fuel

Ueta, Shohei; Aihara, Jun; Sumita, Junya; Shaimerdenov, A.*; Dyussambayev, D.*; Gizatulin, S.*; Chakrov, P.*; Sakaba, Nariaki

no journal, , 

In order to investigate irradiation performance of the newly-designed high temperature gas-cooled reactor (HTGR) fuel for high burnup around 100 GWd/t, a capsule irradiation test has been carried out by WWR-K research reactor in the Institute of Nuclear Physics (INP) of Kazakhstan. A result on evaluation of the fuel integrity based on the fractional release of fission product (FP) released from the fuel during the irradiation and a plan of post-irradiation examination are reported.

Oral presentation

Development on extended burnup fuel technologies for practical high temperature gas-cooled reactors; Collaborative research with Kazakhstan

Ueta, Shohei; Shibata, Taiju; Aihara, Jun; Shaimerdenov, A.*; Dyussambayev, D.*; Takahashi, Masashi*; Kinoshita, Hideaki*; Gizatulin, S.*; Sakaba, Nariaki

no journal, , 

To develop the highly-qualified, mass-produced coated fuel particle in Japan which is supposed to be introduced to small modular commercial high temperature gas-cooled reactors (HTGRs), a dimensional specification of the fuel has been determined to attain three times higher burnup than that of the HTTR (High Temperature Engineering Test Reactor). Fabrication technologies of the fuel have been established in collaboration with Japanese nuclear fuel fabricator. As results on irradiation test and post irradiation examination at the Institute of Nuclear Physics in Kazakhstan via acceptances of two Regular Projects of ISTC (International Science and Technology Center), an excellent performance of the fuel under irradiation has been confirmed. Finally, technologies to extend the burnup for the highly-qualified, mass-produced HTGR fuel have been established first in the world.

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