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JAEA Reports

Validation of repairing method for concrete wall of the JMTR tank-yard building

Sugaya, Naoto; Okada, Yuji; Nishimura, Arashi; Sonobe, Hiroshi; Kimura, Nobuaki; Kimura, Akihiro; Hanawa, Yoshio; Nemoto, Hiroyoshi

JAEA-Testing 2020-004, 67 Pages, 2020/08


In the Japan Materials Testing Reactor (JMTR), the leakage accidents of radioactive waste liquid were occurred from the tanks and pipes of the liquid waste disposal facility in the JMTR tank-yard building in JFY2014. In order to respond to the accident, the tanks and pipes were replaced from JFY2016 to 2019. On the other hand, a lot of cracks were occurred on the concreate wall of the tank-yard building when the frame structure supports were fixed to the concrete wall in the replacement work. Thus, it is necessary to repair the concreate wall of the tank-yard building. Especially, some cracks with swelling (cone-shaped fracture) were raised around some anchor bolts (the post-installed chemical anchor bolts) fixed the frame structure supports. The repairing method for the cone-shaped fracture of the concrete wall is standardized, but there was no reference value of tensile strength for the validation of the post-installed chemical anchor bolts after the repairing method. In this report, the repairing method was selected for the cone-shaped fracture on the concreate wall and the reference value of tensile strength for the validation of the post-installed chemical anchor bolts by this repairing method. The mock-ups for repairing cone-shaped fracture were fabricated by the selected repairing method and the tensile tests of the post-installed chemical anchor bolts were performed. From the results, the validation of the repairing method was obtained in this test and it was obvious the repairing of cone-shaped fracture is preferable method for the concreate wall of the JMTR tank-yard building.

Journal Articles

Feasibility study of tritium recoil barrier for neutron reflectors

Ishitsuka, Etsuo; Sakamoto, Naoki*

Physical Sciences and Technology, 6(2), p.60 - 63, 2019/12

Tritium release into the primary coolant of the research and test reactors during operation had been studied, and it is found that the recoil release from chain reaction of $$^{9}$$Be is dominant. To reduce tritium concentration of the primary coolant, feasibility study of the tritium recoil barrier for the beryllium neutron reflectors was carried out, and the tritium recoils of various materials were calculated by PHITS. From these calculation results, it is clear that the thickness of tritium recoil barrier depends on the material and 20$$sim$$40 $$mu$$m is required for three orders reduction.

JAEA Reports

Setting of EAL in JMTR

Kawamata, Takanori; Onuma, Yuichi; Hanakawa, Hiroki

JAEA-Review 2018-031, 33 Pages, 2019/02


As "Regulations on events to be reported by nuclear disaster prevention manager based on act on special measures concerning nuclear emergency preparedness" was revised and enfored in 2017, Nuclear operator emergency action plan at Oarai Research and Development Institute was reconsidered, and Emergency Action Level (EAL) in JMTR was newly set. In setting the EAL, characteristics of the JMTR and EAL in nuclear power reactors were considerd because the characteristics of research reactors are different in reactor types. This report shows the basic policy and the selection result of the EAL setting in the JMTR.

Journal Articles

Calculation of tritium release from driver fuels into primary coolant of research reactors

Ho, H. Q.; Ishitsuka, Etsuo

Physical Sciences and Technology, 5(2), p.53 - 56, 2019/00

Increasing of tritium concentration in the primary coolant of the research and test reactors during operation had been reported. To check the source for tritium release into the primary coolant during operation of the JMTR and the JRR-3M, the tritium release from the driver fuels was calculated by MCNP6 and PHITS. It is clear that the calculated values of tritium release from fuels are as about 10$$^{7}$$ and 10$$^{6}$$ Bq for the JMTR and JRR-3M, respectively, and that calculated values are about 4 order of magnitude smaller than that of the measured values. These results show that the tritium release from fuels is negligible for both the reactors.

JAEA Reports

Calculations of Tritium Recoil Release from Li and U Impurities in Neutron Reflectors (Joint research)

Ishitsuka, Etsuo; Kenzhina, I.*; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*

JAEA-Technology 2018-010, 33 Pages, 2018/11


As a part of study on the mechanism of tritium release to the primary coolant in research and testing reactors, tritium recoil release rate from Li and U impurities in the neutron reflector made by beryllium, aluminum and graphite were calculated by PHITS code. On the other hand, the tritium production from Li and U impurities in beryllium neutron reflectors for JMTR and JRR-3M were calculated by MCNP6 and ORIGEN2 code. By using both results, the amount of recoiled tritium from beryllium neutron reflectors were estimated. It is clear that the amount of recoiled tritium from Li and U impurities in beryllium neutron reflectors are negligible, and 2 and 5 orders smaller than that from beryllium itself, respectively.

Journal Articles

Evaluation of tritium release curve in primary coolant of research reactors

Ishitsuka, Etsuo; Kenzhina, I. E.*

Physical Sciences and Technology, 4(1), p.27 - 33, 2018/06

Increase of tritium concentration in the primary coolant for the research and testing reactors during reactor operation had been reported. To clarify the tritium sources, a curve of the tritium release rate into the primary coolant for the JMTR and the JRR-3M are evaluated. It is also observed that the amount of released tritium is lower in the case of new beryllium components installation, and increases with the reactor operating cycle. These results show the beryllium components in core strongly affect to the tritium release into the primary coolant. As a result, the tritium release rate is related with produced $$^{6}$$Li by (n,$$alpha$$) reaction from $$^{9}$$Be, and evaluation results of tritium release curve are shown as the dominant source of tritium release into the primary coolant for the JMTR and the JRR-3M are beryllium components. Scattering of the tritium release rate with irradiation time were observed, and this phenomena in the JMTR occurred in earlier time than that of the JRR-3M.

JAEA Reports

Technical review on irradiation tests and post-irradiation examinations in JMTR

Neutron Irradiation and Testing Reactor Center

JAEA-Review 2017-016, 170 Pages, 2017/07


The Japan Materials Testing Reactor (JMTR) has been contributing to various R&D activities in the nuclear research such as the fundamental research of nuclear materials/ fuels, safety research and development of power reactors, radio isotope (RI) production since its beginning of the operation in 1968. Irradiation technologies and post irradiation examination (PIE) technologies are the important factors for irradiation test research. Now, decontamination and new research reactor construction are common issue in the world according to aging. This report outlines any irradiation and PIE technologies developed at JMTR in 40years and for technology succession and development of human resources.

Journal Articles

Influence of temperature histories during reactor startup periods on microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons

Kasahara, Shigeki; Kitsunai, Yuji*; Chimi, Yasuhiro; Chatani, Kazuhiro*; Koshiishi, Masato*; Nishiyama, Yutaka

Journal of Nuclear Materials, 480, p.386 - 392, 2016/11

 Times Cited Count:0 Percentile:100(Materials Science, Multidisciplinary)

This paper addresses influence of two different temperature profiles during startup periods in the Japan Materials Testing Reactor and a boiling water reactor upon microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons to about 1 dpa and 3 dpa. Tensile tests at 290$$^{circ}$$C and Vickers hardness tests at room temperature were carried out, and their microstructures were observed by FEG-TEM. Influence of difference in the temperature profiles was observed obviously in interstitial cluster formation, in particular, growth of Frank loops. The influence was also found certainly in loss of strain hardening capacity and ductility, although the influence on the yield strength and the Vickers hardness was not clearly observed. As a result, Frank loops, which were observed in austenitic stainless steel irradiated at doses of 1 dpa or more, were considered to contribute to deformation of the austenitic stainless steel.

JAEA Reports

Calculation by PHITS code for recoil tritium release rate from beryllium under neutron irradiation (Joint research)

Ishitsuka, Etsuo; Kenzhina, I. E.*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*

JAEA-Technology 2016-022, 35 Pages, 2016/10


As a part of study on the mechanism of tritium release to the primary coolant in research and testing reactors, the calculation methods by PHITS code is studied to evaluate the recoil tritium release rate from beryllium core components. Calculations using neutron and triton sources were compared, and it is clear that the tritium release rates in both cases show similar values. However, the calculation speed for the triton source cases is two orders faster than that for the neutron source case. It is also clear that the calculation up to history number per unit volume of 2$$times$$10$$^{4}$$ (cm$$^{-3}$$) is necessary to determine the recoil tritium release rate of two effective digits precision. Furthermore, the relationship between the beryllium shape and recoil tritium release rate using the triton sources was studied. Recoil tritium release rate showed linear relation to the surface area per volume of beryllium, and the recoil tritium release rate showed about half of the conventional equation value.

JAEA Reports

Performance tests of radiation detectors for inspection of $$^{99}$$Mo/$$^{99m}$$Tc solution, 1

Suzuki, Yumi*; Nakano, Hiroko; Suzuki, Yoshitaka; Ishida, Takuya; Shibata, Akira; Kato, Yoshiaki; Kawamata, Kazuo; Tsuchiya, Kunihiko

JAEA-Technology 2015-031, 58 Pages, 2015/11


Technetium-99m ($$^{99m}$$Tc) is one of the most commonly used radioisotopes in the field of nuclear medicine. In the Japan Atomic Energy Agency (JAEA), the research and development (R&D) have been carried out for production of molybdenum-99 ($$^{99}$$Mo) by (n, $$gamma$$) method, a parent nuclide of $$^{99m}$$Tc, with the Japan Material Testing Reactor (JMTR). On the other hand, the new project as "Domestic Production of Medical Radioisotope (Technetium preparation) in Japan" was adopted in the Tsukuba International Strategic Zone on October, 2013 and the demonstration tests will be planned for the domestic production of $$^{99}$$Mo/$$^{99m}$$Tc with the JMTR. Thus, new facilities and analysis devices were equipped in the JMTR hot laboratory in 2014 as the part of this project. As the part of the analytical device equipment, the $$gamma$$-TLC analyzer and the radiation detector connected with the High Performance Liquid Chromatography (HPLC) were installed for quality inspection of the $$^{99}$$Mo/$$^{99m}$$Tc solution and the extracted $$^{99m}$$Tc solution in the JMTR hot laboratory. The performance tests of these devices such as detection sensitivity, resolution, linearity and selectivity of energy range were carried out with $$^{137}$$Cs and $$^{152}$$Eu as alternative radionuclides of $$^{99}$$Mo and $$^{99m}$$Tc, respectively. In the results, bright prospects were obtained concerning the quality inspection of the $$^{99}$$Mo/$$^{99m}$$Tc and $$^{99m}$$Tc solutions using these devices. This report describes the results of those performance tests.

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


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

Zirconia membrane ECP electrode for water chemistry experiments in the JMTR

Hanawa, Satoshi; Hata, Kuniki; Shibata, Akira; Chimi, Yasuhiro; Kasahara, Shigeki; Tsutsui, Nobuyuki*; Iwase, Akihiro*; Nishiyama, Yutaka

Proceedings of 2014 Nuclear Plant Chemistry Conference (NPC 2014) (USB Flash Drive), 9 Pages, 2014/10

no abstracts in English

Journal Articles

Present status of in-pile IASCC growth tests at JMTR

Kaji, Yoshiyuki; Ugachi, Hirokazu; Nakano, Junichi*; Matsui, Yoshinori; Kawamata, Kazuo; Tsukada, Takashi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*

HPR-364, Vol.1 (CD-ROM), 10 Pages, 2005/10

Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors (LWRs) for a long period. In the research field of IASCC, mainly PIEs for irradiated materials have been carried out, because there are many difficulties on SCC tests under neutron irradiation. Hence as a part of the key techniques for in-pile SCC tests, we have embarked on a development of the test technique to obtain information concerning effects of applied stress level, water chemistry, irradiation conditions, etc. In this paper, we describe the developed several techniques, especially control of loading on specimens, monitoring technique of crack propagation and so on, and the present status of in-pile IASCC growth tests using pre-irradiated materials at JMTR.

Journal Articles

Study of irradiation effects on fuels and materials in JMTR hot laboratory

Omi, Masao

Nippon Genshiryoku Gakkai Kaku Nenryo Bukai Dai-20-kai "Kaku Nenryo, Kaki Semina" Kogi Tekisuto, 8 Pages, 2005/07

no abstracts in English

JAEA Reports

An Irradiation experiment for qualification of insulating coating

Nakamichi, Masaru*; Kawamura, Hiroshi

JAERI-Research 2005-015, 35 Pages, 2005/06


no abstracts in English

JAEA Reports

Low temperature tritium release experiment from lithium titanete breeder material

Tsuchiya, Kunihiko; Kawamura, Hiroshi; Nakamichi, Masaru*; Sagawa, Hisashi

JAERI-Tech 2005-013, 56 Pages, 2005/03


no abstracts in English

Journal Articles

Effects of helium production and heat treatment on neutron irradiation hardening of F82H steels irradiated with neutrons

Wakai, Eiichi; Taguchi, Tomitsugu; Yamamoto, Toshio*; Tomita, Hideki*; Takada, Fumiki; Jitsukawa, Shiro

Materials Transactions, 46(3), p.481 - 486, 2005/03

 Times Cited Count:5 Percentile:53.13(Materials Science, Multidisciplinary)

no abstracts in English

JAEA Reports

Examination on detaching method of an irradiated mock-up with Li$$_{2}$$TiO$$_{3}$$ pebble bed from the core of JMTR

Ikeshima, Yoshiaki; Ishida, Takuya*; Tsuchiya, Kunihiko; Tomita, Kenji; Ebisawa, Hiroyuki; Magome, Hirokatsu; Nakamichi, Masaru*; Kitajima, Toshio; Kawamura, Hiroshi

JAERI-Tech 2005-005, 37 Pages, 2005/02


no abstracts in English

JAEA Reports

A Preliminary irradiation test of the new ceramics materials for the innovative basic research on high-temperature engineering

Baba, Shinichi; Yamaji, Masatoshi*; Shibata, Taiju; Ishihara, Masahiro; Sawa, Kazuhiro

JAERI-Tech 2005-002, 83 Pages, 2005/02


The innovative basic research concerns the association of basic and technology with high temperature irradiation environments. The direction of research involves novel and bold challenges both in nuclear and non-nuclear fields. The innovative basic research on high-temperature engineering is one of the key subjects using the HTTR as well as other research reactors. The aim of this paper is to describe the preliminary irradiated test conditions such as irradiated fluence, irradiated temperature and installed specimen on the capsules, and also to show the specifications and performances of the apparatus for Post/Ante Irradoation Experiment.

JAEA Reports

Annual report of JMTR, No.18, FY2003; April 1, 2003 - March 31, 2004

Department of JMTR

JAERI-Review 2004-029, 100 Pages, 2005/01


no abstracts in English

303 (Records 1-20 displayed on this page)