Significant leakage of the primary containment vessel (PCV) occurred on March 15, 2011 at Unit 2 of Fukushima Dai-ichi Nuclear Power Station, causing land contamination over a large area. From the radiation dose rate map of Unit 2 and the temperature trend of the PCV, degradation of the PCV top head flange gasket was identified as the cause of the leakage. The design set point of the rapture disc and operability of the valves in the hardened containment venting system were further identified as factors contributing to the leakage. Based on the lessons from the leakage, the gasket material was improved to have greater heat resistance in a steam environment, the PCV cooling system was diversified to improve reliability, and the operability and operation methods of the PCV venting were improved for greater robustness of the PCV in severe accidents.
A high-temperature engineering test reactor (HTTR) has a reactivity control system which is accompanied with a reserved shutdown system (RSS). The RSS consists of B4C/C pellets, a guide tube, an electric plug, a motor which contains a brake and reducer, and so on. In accidents when the control rods cannot be inserted, the electric plug is pulled out by the motor and the B4C/C pellets fall into the core by gravity. It was revealed that the motor in the RSS drive mechanism did not work as the result of pre-start-up checks as described in the following: (1) The oil which separated from the grease of the motor reducer flowed down from the gap of the oil seal, (2) the separated oil penetrated into the brake, (3) the penetrated oil mixed with the abrasive powder released from the brake disk, and finally, (4) the adhesive mixture blocked the rotation of the motor. A new evaluation method was proposed to detect signs of the motor sticking. Through the overhaul inspection of all RSS drive mechanisms of an HTTR, it was revealed that the proposed method was effective to evaluate the integrity of the RSS drive mechanism.
One of the most important issues of the current PRA methodology is the precise modeling of dynamic changes such as state transitions among several states including fault(s) or maintenance of the nuclear facility, safety-related systems or components by fault-tree analysis and event-tree analysis. Moreover, though safety-related systems are usually in the stand-by state during normal operating conditions of a nuclear power plant, modeling of the dynamic changes in safety functions, along with changes in component failure rates due to the aging effect in the stand-by state or continuous/intermittent effects originating from external hazards, is also carried out under the same situation. On the basis of the background described above, the authors proposed a reliability analysis methodology of using the Markov state transition model applied to the digital reactor protection system of an ABWR plant, and demonstrated the applicability of the developed methodology using the component failure modes discussed in DIGREL, the task group of WGRisk belonging to OECD/NEA/CSNI. These studies showed that the PRA methodology including the state transition model can consider state transitions of components and time-dependent changes in component failure rates, and the relationship between this methodology and minimal cut sets for calculating the core damage frequency was also clarified.
Three nuclear reactors at Fukushima Daini Nuclear Power Station lost all their ultimate heat sinks owing to damage from the tsunami caused by the Great East Japan Earthquake on March 11, 2011. Water was injected into the reactors by alternate measures, damaged cooling systems were restored with promptly supplied substitute materials, and all the reactors were brought to a cold shutdown state within four days. Lessons learned from this experience were identified to improve emergency management, especially in the areas of strategic response planning, logistics, and functions supporting response activities continuing over a long period. It was found that continuous planning activities reflecting information from plant parameters and response action results were important, and that relevant functions in emergency response organizations should be integrated. Logistics were handled successfully but many difficulties were experienced. Therefore, their functions should be clearly established and improved by emergency response organizations. Supporting emergency responders in the aspects of their physical and mental conditions was important for sustaining continuous response. As a platform for improvement, the concept of the Incident Command System was applied for the first time to a nuclear emergency management system, with specific improvement ideas such as a phased approach in response planning and common operation pictures.
The mechanical analysis code MACBECE2014 has been developed at the Japan Atomic Energy Agency (JAEA) to make realistic simulations of the physical integrity of the near field for performance assessment of the geological disposal of TRU waste in Japan. The MACBECE2014 code can be used to evaluate long-term changes in the mechanical behavior of the near field and any subsequent changes in the permeability of engineering barrier components, including crack formation in cementitious materials caused by expansion due to metal corrosion. Cracks in cementitious materials are likely to channel the flow of groundwater and so the represent preferred flow paths of any released radionuclides. Mechanical analysis was conducted using the MACBECE2014 code to investigate the concept of the TRU waste disposal system described in JAEA’s Second Progress TRU Report. Simulated results of a disposal system with a bentonite buffer demonstrated that the low permeability of the engineering barrier system could be maintained for long time periods because the physical integrity of the bentonite buffer remained intact even if cracks in the cementitious components had formed locally. Simulated results of the disposal system with a concrete backfill instead of a bentonite buffer showed that crack formation leads to a significant increase in the permeability of the engineering barrier system.
A non-destructive assay system using the fast neutron direct interrogation method has been designed and developed to be put into practical use for the determination of the uranium (235U) mass contained in actual uranium-contaminated waste drums. The method is capable of measuring the fissile mass in a drum by counting the number of fission neutrons resulting from nuclear fission reactions between the fissile materials contained in a drum and thermal neutrons generated by 14 MeV fast neutrons irradiated from outside the drum. A performance test employing simulated metal waste drums demonstrated that a natural uranium mass of as low as approximately 10 g could be detected within an error of ±20% regardless of the distribution of uranium samples in the drum, and the total number of fission neutrons was proportional to the 235U mass. A demonstration test employing actual waste drums could determine the uranium mass by using a newly developed correction method for deriving the fissile mass in a drum. It has been proved by the experimental validation tests that the assay system equipped with the correction method is very useful for the accountancy of waste drums.