The purpose of this research is to establish a stmctural reliability evaluation method for aged nuclear components under seismic events from the viewpoint of long-term use of existing light water reactor nuclear power plants. For this purpose, we developed a piping failure probability evaluation code "PASCAL-SC" and seismic hazard evaluation code "SHEAT-FM" which is reported separately. In PASCAL-SC, aging such as SCC (Stress Corrosion Cracking) and fatigue crack extension by seismic load is considered. In this paper, a reliability evaluation procedure for aged piping under seismic events by combination of these codes is described. We evaluate the reliability of a welded joint of PLR (Primary Loop Recirculation system) piping using a BWR model plant under seismic events.
Melting-solidified waste form is expected to act as an engineered barrier due to the stabilization of radionuclides and hazardous elements. Solidified products are categorized to metal and slag, which originate from metallic and non-metallic components, respectively. Static leach tests were performed for SiO2-CaO-Al2O3 slag specimens in the presence of cement at 90°C to investigate the dissolution behavior of the slag under repository conditions. The dissolution rate of slag was constant during the leaching up to 144 days. Dissolution of slag was followed by precipitation of calcium silicate hydrates, which is expected to depress the concentration of Si dissolved silica in solution and to maintain the solution conditions far from the saturation with the slag. The far-from-saturation state is likely to be a cause of the constant dissolution rate.
Monte Carlo seminars have been held at Japan Atomic Energy Institute. There are 1) Monte Carlo fundamental theory, 2) sub-criticality seminar with Monte Carlo method for nuclear fuel cycle facilities, 3) shielding safety analysis seminar with Monte Carlo method, 4) estimation method of lower weight bound for neutron deep penetration problem with Monte Carlo method, 5) shielding safety analysis at high energy with MCNPX code, 6) streaming safety analysis seminar with Monte Carlo method, 7) skyshine analysis seminar with Monte Carlo method, 8) dosimetry seminar with Monte Carlo method. In the fundamental theory seminar, new estimation method for lower weight bound or importance are introduced. In the sub-criticality seminar, eleven benchmark experiment problems conducted by the authors are introduced in order to confirm the reliability of calculation method.
The MK-III upgrading project was completed in the experimental fast reactor JOYO to increase irradiation capability for irradiation tests. The performance tests were carried out from June 2003 as the last phase of MK-III modification work. During the performance tests, the reactor power was raised step by step, while confirming the nuclear and thermal characteristics of MK-III core and the heat removal capability of the intermediate heat exchanger (IHX) and the dump heat exchanger (DHX). All performance tests were successfully carried out and it was confirmed that the performance of JOYO MK-III plant satisfied the design requirement. A pre-use inspection pass certificate for JOYO MK-III was granted from Ministry of Education, Culture, Sports, Science and Technology in 27th November 2003 and the MK-III modification work was completed. This report shows the results of the performance tests of JOYO MK-III.