Etsuo Ishitsuka

Feasibility study of large-scale production of iodine-125 at the high temperature engineering test reactor

The feasibility of a large-scale iodine-125 production from natural xenon gas at high-temperature gas-cooled reactors (HTGRs) was investigated. A high-temperature engineering test reactor (HTTR), which is located in Japan at Oarai-machi Research and Development Center, was used as a reference HTGR reactor in this study. First, a computer code based on a Runge-Kutta method was developed to calculate the quantities of isotopes arising from the neutron irradiation of natural xenon gas target. This code was verified with a good agreement with a reference result. Next, optimization of irradiation planning was carried out. As results, with 4 days of irradiation and 8 days of decay, the 125I production could be maximized and the 126I contamination was within an acceptable level. The preliminary design of irradiation channels at the HTTR was also optimized. The case with 3 irradiation channels and 20-cm diameter was determined as the optimal design, which could produce approximately 1.8 × 105GBq/y of 125I production.

Status of development on 99 Mo production technologies in JMTR

The Japan Materials Testing Reactor (JMTR) is now under refurbishment, and the operation of the new JMTR will be started in FY 2011. The new JMTR has a plan to produce 99Mo, which is the parent nuclide of 99mTc, and two 99Mo production technologies have been developed; the one is a solid irradiation method, and the other is a solution irradiation method. In the solid irradiation method, it was found that JMTR can provide about 20% of the 99Mo imported into Japan. In the solution irradiation method, the fundamental characteristics of the aqueous molybdate solutions selected as the irradiation target were cleared by the gamma-ray irradiation test. Aiming at the domestic production of 99Mo in Japan, the development of 99Mo production technologies in JMTR has been continued.

Development of Tritium Release Apparatus Using Pulse Mode Heating

Tritium release apparatus with the function of pulse mode heating was developed by using the infrared ray furnace to demonstrate the pulse mode heating of tritium breeder blanket for the fusion reactor. This apparatus was installed in the glove box of the beryllium PIE facility that constructed the hot-laboratory of Japan Materials Testing Reactor. The performance of this apparatus is that the minimum time of rapid heating up to 1015 {degree}C is about 119 s and maximum heating rate reached at 1000 {degree}C/min. The maximum temperature depends on the crucible materials because of the differences for infrared ray absorption. The conversion efficiency of the gaseous water by ceramic electrolysis cell is above 99.99%. The pulse mode heating of the tritium breeder and neutron multiplier materials of the blanket could be demonstrate by using this apparatus. 7 refs., 7 figs., 1 tab.