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Tc generators loading processTanimoto, Masataka; Amaya, D.*; Aoyama, Masashi; Kimura, Akihiro; Izumo, Hironobu; Tsuchiya, Kunihiko
JAEA-Review 2011-012, 13 Pages, 2011/06
JAEA-Review-2011-012.pdf:2.1MB
Tc is most commonly used as a radiopharmaceutical in the field of nuclear medicine, accounting for more than 80% of all diagnostic nuclear medicine procedure. The Tc is obtained from 
Mo, which is produced by fission of 
U ((n, f) method) and the neutron capture (n, 
) method using the 
Mo target. However, a supplying of Mo is only depends on imports from any other countries, so JAEA find a way out at domestic production of a part of Mo that (n, ) method in cooperation with the industrial circles. On the other hand, INVAP has been working in the supply of Mo production facilities using LEU. This report provides descriptions the detail technical aspects related to the facility and operations for loading Tc generator. These key issues and technical provided in this report is believed to be useful for developing and updating them.
Mo production processTanimoto, Masataka; Amaya, D.*; Aoyama, Masashi; Kimura, Akihiro; Izumo, Hironobu; Tsuchiya, Kunihiko
JAEA-Review 2011-002, 26 Pages, 2011/03
JAEA-Review-2011-002.pdf:1.46MB
Recently, worldwide demand of Mo became rises. However the availability and supply of Mo for the manufacturing of generators has been a matter of concern. Concern arose from several factors including, amongst others, the shutdown of some nuclear reactors at Canada (NRU, etc.), uncertainty of reliable operating condition for radioisotope production and difficulties in the availability of highly enriched U (HEU) target material used in the majority of the production facilities. As countermeasure for this issue, the HEU is not used but Mo production from low enriched U (LEU) is performed. This production process was developed in Argentina by the Argentine Atomic Energy Commission (CNEA). In the last ten years, INVAP has been working in the supply of Mo production facilities using LEU. This report provides descriptions for the detail technical aspects related to a Mo production system using irradiated LEU targets.
Ishitsuka, Etsuo; Kitagishi, Shigeru; Aoyama, Masashi; Kawamata, Kazuo; Nagao, Yoshiharu; Ishihara, Masahiro; Kawamura, Hiroshi
Proceedings of 1st Asian Symposium on Material Testing Reactors (ASMTR 2011), p.111 - 115, 2011/02
Tanimoto, Masataka; Aoyama, Masashi; Kitagishi, Shigeru; Shibata, Akira; Saito, Takashi; Nakamura, Jinichi; Tsuchiya, Kunihiko
Proceedings of 1st Asian Symposium on Material Testing Reactors (ASMTR 2011), p.62 - 70, 2011/02
The new JMTR is expected to contribute to many fields: the lifetime extension of LWRs (aging management of LWRs, development of next generation LWRs, etc.), the expansion of industry use (production of the medical radioisotope 
Tc, etc.) and the progress of science and technology (namely, basic research on nuclear energy). To meet a wide range of users needs, new irradiation technologies with advanced techniques have been developed. In this paper, status of the development of new measuring instruments are introduced which for neutron and irradiation tests in JMTR such as multi-paired thermocouple, rod inner pressure gauge, Self-Powered Neutron Detector (SPND) and Self-Powered Gamma Detector (SPGD).
Shibata, Akira; Kimura, Tadashi; Nagata, Hiroshi; Aoyama, Masashi; Kanno, Masaru; Omi, Masao
JAEA-Testing 2010-003, 22 Pages, 2010/11
JAEA-Testing-2010-003.pdf:8.82MB
Type 316 stainless steels (SSs) were used for tube material of the Oarai water loop No.2 (OWL-2) in the reactor. But data of highly irradiated Type 316 SSs has been insufficient since OWL-2 was installed. Therefore surveillance tests of type 316 SSs which were irradiated up to 3.4 
10
n/m
in fast neutron fluence (
1 MeV) were performed. But type 316 SSs were widely used in JMTR, then additional data of type 316 SSs irradiated higher was required. Therefore PIEs of type 316 SSs surveillance specimens which were irradiated up to 1.0 10
n/m in fast neutron fluence were performed and reported in this paper. Tendency of results has good agreement with results of 10
-10 n/m in fast neutron fluence. More than 37 % in total elongation was confirmed in all test conditions. It is confirmed that type 316 SS irradiated up to 1.0 10 n/m in fast neutron fluence has enough ductility as structure material.
Kitagishi, Shigeru; Aoyama, Masashi; Tobita, Masahiro; Inaba, Yoshitomo; Yamaura, Takayuki
no journal, ,
For the up-graded use and the lifetime extension of LWRs, the JMTR has a plan of power ramping tests on new design LWR fuels. The purpose of the power ramping tests is to evaluate the fuel behavior during power ramping condition using a capsule inserted into the OSF-1. The design and fabrication of the natural convection capsule used in the power ramping tests has been carried out based on BOCA, which had experience on power ramping tests of 88 BWR fuels. The outer cladding diameter of 88 BWR fuels is 12 mm. On the other hand, for a high power, the outer cladding diameter of the new design LWR fuel is changed to 9.5 mm. And, there is potential for transition from nucleate boiling to film boiling because the heat flux of the peak linear power 600 W/cm in the tests approaches the critical heat flux. In this study, prior to the irradiation tests of the fuels, the out-of pile test facility, which had a heater pin instead of a test fuel pin, was designed and fabricated to simulate the capsules used in power ramping tests, and thermal-hydraulic tests were carried out using the out-of pile test facility. In addition, in order to investigate the thermal-hydraulic behavior in the capsule, the out-of pile tests were simulated numerically using the improved analysis code ACE-3D. From these experimental and analytical results, it was found that the power ramping tests with fuels with their outer cladding diameters of 9.5 mm are able to be realized up to linear power 600 W/cm in JMTR.
Kitagishi, Shigeru; Isozaki, Futoshi; Takita, Kenji; Aoyama, Masashi; Matsui, Yoshinori
no journal, ,
As one of the production of radioisotopes, JMTR has a plan to produce Mo which is a parent nuclide of Tc. The new hydraulic-rabbit-irradiation-facility is carried out to be set up in the preparation of the Mo production facility. For the expansion of use, the new facility was designed to be able to irradiate rabbits more than it used to be. However, it takes more than about ten days to assemble the rabbit. For this reason, it is required shortening of the assembly period. The improved type of rabbit, which was developed to shorten the assembly period in JRR-3, was focused. And, out-pile tests were carried out for research of the fabrication of the improved type rabbit and the endurance performance under the usage environment in JMTR. In these tests, first, the assembly device for the improved type of rabbit was designed and fabricated, and the fabrication test was carried out using the assembly device. Next, the endurance performance tests were carried out using mock-up rabbits. From the results of these tests, there were bright prospects of the manufacturability of the improved type of rabbit had good sealing performance and utilizing in the irradiation test of JMTR.