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Shaimerdenov, A.*; Gizatulin, Sh.*; Sairanbayev, D.*; Bugybay, Zh.*; Silnyagin, P.*; Akhanov, A.*; Fuyushima, Takumi; Hirota, Noriaki; Tsuchiya, Kunihiko
Nuclear Instruments and Methods in Physics Research B, 548, p.165235_1 - 165235_6, 2024/03
Times Cited Count：0Compared to conditions in other types of installations, cable insulation in nuclear reactors is exposed to mixed conditions (high temperatures, radiation, pressure, humidity, aggressive environments) and at the same time they must maintain their performance characteristics for a long time (about 40-50 years). As a result of irradiation to such conditions, the electrical properties of the cable insulation are degraded, which leads to an increase in current loss. This is because the charge is induced by radiation into the insulator. At the WWR-K reactor, studies were started on the radiation resistance of signal cables with two types of mineral insulation (MgO and Al
O
). As part of these studies, new experimental data will be obtained on the behavior of signal cables with mineral insulation of two types in mixed operating conditions (radiation field and high temperature). It is planned to accumulate fluence of fast neutrons 
10
cm
in cables. The irradiation temperature will be (500 
50)
C). The study of the degradation of the electrical properties of the insulation of signal cables will be carried out in real time (in-situ). For this, a special design of the experimental device and a technique for in-reactor measurement of electrical characteristics were developed. This paper presents a sketch of the capsule design, the results of complex calculations for the development of the capsule design, the expected neutron fluences, the dpa in steel, the technique for in-reactor measurement of electrical characteristics, and a work plan for the future indicating the expected results. The cable irradiation time until the target neutron fluence is reached will be about 100 effective days. This research is funded by the International Scientific-Technical Center.
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; 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.
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.
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.
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.
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.
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 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.
Chekushina, L.*; Dyussambayev, D.*; Shaimerdenov, A.*; Tsuchiya, Kunihiko; Takeuchi, Tomoaki; Kawamura, Hiroshi; Kulsartov, T.*
Journal of Nuclear Materials, 452(1-3), p.41 - 45, 2014/09
Times Cited Count：6 Percentile：26.79(Materials Science, Multidisciplinary)no abstracts in English
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.
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.
