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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

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

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

Preliminary test for Mo recycling system in $$^{99}$$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 $$^{99}$$Mo, the parent nuclide of $$^{99m}$$Tc by the (n,$$gamma$$) method using JMTR. The (n,$$gamma$$) method has an advantage of easy manufacturing process and low radioactive wastes generation. However, the low radioactivity concentration of $$^{99m}$$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.

Journal Articles

Evaluation of selected grades of beryllium metal as reflector materials for extended lifetime performance

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

Journal Articles

Development of beryllium material for reflector lifetime expansion

Dorn, C. K.*; Tsuchiya, Kunihiko; Hatano, Yuji*; Chakrov, P.*; Kodama, Mitsuo*; Kawamura, Hiroshi

Proceedings of 5th International Symposium on Material Testing Reactors (ISMTR-5) (Internet), 9 Pages, 2012/10

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 and interaction between pure water and these beryllium grades are accumulated under un-irradiated. Additionally, irradiation tests have been prepared and development of PIE technologies has been performed. In this paper, the results of various properties and irradiation test plan for lifetime expansion of beryllium are described for material testing reactors.

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.

Journal Articles

Status of $$^{99}$$Mo-$$^{99m}$$Tc production development by (n,$$gamma$$) reaction

Tsuchiya, 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 $$^{99}$$Mo by (n,$$gamma$$) method, a parent nuclide of $$^{99m}$$Tc. In case of Japan, the supplying of $$^{99}$$Mo depends only on imports from foreign countries, the R&D on production method of $$^{99}$$Mo-$$^{99m}$$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$$_{3}$$ pellets, (2) Separation and concentration of $$^{99m}$$Tc by the solvent extraction from Mo solution, (3) Examination of $$^{99m}$$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 $$^{99}$$Mo-$$^{99m}$$Tc.

Journal Articles

Status of material development for lifetime expansion of beryllium reflector

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.

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.

JAEA Reports

Retesting of $$^{99}$$Mo adsorption and $$^{rm 99m}$$Tc elution characteristics of molybdenum adsorbents for (n,$$gamma$$) method; Joint experiment report on irradiation technology of RI production (STC No.2$$-$$II) (Joint research)

Kimura, Akihiro; Izumo, Hironobu; Tsuchiya, Kunihiko; Hori, Naohiko; Ishihara, Masahiro; Bannykh, V.*; Gluschenko, N.*; Chakrova, Y.*; Chakrov, P.*

JAEA-Testing 2010-002, 20 Pages, 2010/08

JAEA-Testing-2010-002.pdf:2.92MB

JMTR has a plan to produce $$^{99}$$Mo, which is the parent nuclide of radiopharmaceutical $$^{rm rm 99m}$$Tc, by (n,$$gamma$$) method. The cooperation experiments for $$^{99}$$Mo adsorption and $$^{rm 99m}$$Tc elution with the Poly-Zirconium Compound (PZC) and the Molybdate Zirconium Gel (Zr-gel) methods were carried out at Kazakhstan National Nuclear Energy Center (NNC) in October, 2009. The $$^{99}$$Mo adsorption capability was the same level as reference data, however the $$^{rm 99m}$$Tc elution capability with PZC was lower than reference data in this test. Therefore, re-experiments of $$^{99}$$Mo adsorption and $$^{rm 99m}$$Tc elution with both methods were carried out at NNC. As a result, the $$^{99}$$Mo adsorption and $$^{rm 99m}$$Tc elution capabilities were obtained as the same levels as reference data. Additionally, $$^{rm 99m}$$Tc solution was high purity by the elution method connected with alumina column.

JAEA Reports

$$^{99}$$Mo adsorption and $$^{rm 99m}$$Tc elution characteristics of molybdenum adsorbents for (n,$$gamma$$); Method-joint experiment report on irradiation technology of RI production (STC No. 2-II) (Joint research)

Kimura, Akihiro; Izumo, Hironobu; Tsuchiya, Kunihiko; Hori, Naohiko; Ishihara, Masahiro; Bannykh, V.*; Gluschenko, N.*; Chakrova, Y.*; Chakrov, P.*

JAEA-Technology 2009-075, 23 Pages, 2010/02

JAEA-Technology-2009-075.pdf:7.41MB

Japan Materials Testing Reactor (JMTR) of the Japan Atomic Energy Agency (JAEA) has a plan to produce $$^{99}$$Mo, which is the parent nuclide of radiopharmaceutical $$^{rm 99m}$$Tc, by (n,$$gamma$$) method. The $$^{99}$$Mo adsorption and $$^{rm 99m}$$Tc elution characteristics of molybdenum adsorbents should be evaluated since the specific activity of $$^{99}$$Mo obtained by (n,$$gamma$$) method is low. Therefore, $$^{99}$$Mo adsorption and $$^{rm 99m}$$Tc elution tests with molybdenum adsorbents for the (n,$$gamma$$) method such as poly-zirconium compound (PZC) and molybdate zirconium gel were carried out under cooperation with the Kazakhstan National Nuclear Energy Center (NNC). As a result, the $$^{99}$$Mo adsorption performance of the adsorbents was the same level as conventional data, whereas the $$^{rm 99m}$$Tc elution performance of the adsorbents was lower than conventional data. The $$^{99}$$Mo adsorption and $$^{rm 99m}$$Tc elution performance will be investigated again in future.

Oral presentation

Status of beryllium reflector development in JMTR

Dorn, C. K.*; Tsuchiya, Kunihiko; Hatano, Yuji*; Chakrov, P.*; Kodama, Mitsuhiro*; Kawamura, Hiroshi

no journal, , 

no abstracts in English

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.

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