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An accelerator-driven deuterium-lithium (d-Li) stripping reaction-type neutron source, such as the International Fusion Materials Irradiation Facility (IFMIF) planned by the International Energy Agency is recognized as one of the most promising facility to obtain test environments of high-energy neutrons for fusion reactor materials development. The limitation on the available irradiation volume of the irradiation facility requires the development of the small specimen test techniques (SSTT). Application of SSTT to evaluate the degradation of various components in the light water reactor for the life extension is expected to be also quite beneficial.A remote-controlled testing machine for the Small Punch (SP) and miniaturized tensile tests was developed at the hot laboratory of the Japan Materials Testing Reactor (JMTR). The machine is designed for testing at temperatures ranging between 93 and 1, 123K to evaluate the temperature dependence of the strength of materials including the embrittlement at low temperatures and the softening at elevated temperatures. The tests are performed in a vacuum or in an inert gas environment.The machine has been installed in a hot cell and is being used for the round robin test program of the SP test method. The round robin test program is planned to identify the capability of the test method and to establish a standard test procedure. The configuration and the specifications of the test machine are introduced and the results of the SP tests are also shown.

A high temperature in-pile gas loop OGL-1 has been developed, constructed and operated successfully in the Japan Materials Testing Reactor (JMTR) at the Oarai Research Establishment of Japan Atomic Energy Research Institute. The loop was developed as a testing facility for the future multi-purpose high temperature gas cooled reactor (VHTR), specifically designed to make irradiation tests of its nuclear fuel and graphite under the helium flow condition of 1, 000°C.Its design and manufacturing had a great number of developmental factors as there was few experience for the systems, components and structures, operating under such high temperature conditions. Thus OGL-1 is not only an irradiation facility but also a pilot plant of VHTR.After 7 years of many laborlous developments OGL-1 was completed in March 1977, and in line with the original objectives it has achieved a 1, 000°C continuous operation inviting the successful operation at the present time.This report first delineates the general outline of the OGL-1 and makes reference to the R & D actually employed in developing OGL-1. Then safety and structural designs incorporating developmental test results are represented in details. Finally successful performance tests of OGL-1 which enable irradiation tests of fuel specimen under the helium flow conditions of 1, 000°C, are described briefly.

During normal operation of Japan Materials Testing Reactor (JMTR), radioiodines are always detected in the primary cooling water in low levels. In order to identify the main source of the radioiodines, the data obtained during steady-state operation were investigated and uranium contents in main structural materials of JMTR core were determined by neutron activation analysis. It was found that the concentration of the radioiodines had varied in accordance with the exchange of the beryllium reflector frame in the core. The analytical results of the main structural materials showed that the uranium content in beryllium materials was 10-42ppm. The radioiodine concentration in the water calculated from the uranium content agreed with the measured concentration with a limited accuracy. These results suggest that the radioiodines in the reactor water were released mainly from uranium contained in beryllium materials as an impurity.

Using four kinds of fuel rods, i.e. one 8×8 standard type fuel rod, two kinds of high performance fuel rods (with Cu-barrier and with Zr-liner) and one fuel rod filled with He/Xe mixed gas, the fuel centerline ternperatures measurements were conducted at reactor scram of the Japan Materials Testing Reactor (JMTR) by thermocouples instrumented to each fuel rod.From the temperature response of these thermocouples at reactor scram, we estimated the thermal time-constants of the fuel rods and investigated whether the fuel centerline temperature responses of LWR fuel rods with different designs are different.It was concluded that the thermal resistance of the fuel rod with the Zr-liner is higher than that of the fuel rod with the Cu-barrier where the inside surface roughness of the cladding with the Zr-liner is 6.6μm, because the thermal time-constant of the fuel rod with the Zr-liner is greater than that of the fuel rod with the Cu-barrier.