Photon energy spectra were measured above the operating floor of unit 3 reactor at the Fukushima Daiichi Nuclear Power Station by using a CdZnTe semiconductor spectrometer. The spectrometer was installed in a lead collimator to measure the photons from the area directly below the detector. The collimator and spectrometer were lifted up by a huge crane and set above the operating floor. The photon spectra were derived by unfolding the pulse height spectra measured using the spectrometer. The response function of the spectrometer was calculated with the MCNP-4C code and was used as an input parameter of the unfolding code MAXED. It was found from the photon energy spectra that low-energy photons with energy below 0.4 MeV were dominant above the operating floor. These spectra are fundamental data for evaluating the dose reduction effect by setting up shields on the operating floor.

　A seismic probabilistic risk assessment (PRA) for fast breeder reactors (FBRs) has been carried out to confirm that the seismic safety is equivalent to that of light-water reactors (LWRs). FBR plants consist of components with thin walls, so the seismic responses tend to be large in comparison with LWRs. The seismic response of the reactor structure of FBRs is caused by seismic reactivity. The group motion of fuel subassemblies is one typical seismic response. Thus, much attention has been paid to the reactivity insertion mechanism and its consequences during earthquakes beyond the design basis ground motion (DBGM) condition. Continued seismic reactivity insertion may cause fuel melting. To prevent a core disruptive accident (CDA), it is necessary to terminate the expansion of the fuel melting zone by the insertion of control rods. The behaviors of both the fuel melting and the scram delay depend on the seismic exiting acceleration, so the limit to prevent a CDA is defined as the critical acceleration for which the scram time is equivalent to the limit time to prevent fuel pin failure from the viewpoint of the fuel melting behavior. To clarify the critical acceleration and to clarify the uncertainty in the evaluation of the fragility, the evaluation method is studied in terms of the seismic response of the reactor structure, the neutronic characteristics and thermal hydraulic characteristics.

 It is important to accurately estimate the intake quantity for reliable internal exposure assessments. The intake quantity has been estimated by using the least-squares method. However, to use the least-squares method, the number of radioactivity measurements must be more than the number of intakes. To remedy this restriction, this study suggests an estimation method using singular value decomposition that is available regardless of the relation between the numbers of measurements and intakes. Moreover, this study introduces a procedure to calculate the intake quantity from the measurements with uncertainty.

 Cavitation is a highly important issue in various fluid machineries. In the design of an advanced loop-type sodium-cooled fast reactor in Japan, vortex cavitation is also a significant issue for the integrity of the reactor structure. Thus, an evaluation method for vortex cavitation is required. In this study, vortex cavitation at a single suction pipe inlet was studied under several different viscosity conditions including its transient behavior. The intermittent occurrence behaviors of vortex cavitation were grasped by visualization measurements. The experimental results showed that the influence of the kinematic viscosity was obvious under a high kinematic viscosity. However, the influence became smaller with decreasing kinematic viscosity. From these results, the non-dimensional circulation, which was defined as the ratio of the local circulation to the kinematic viscosity, was deduced as an evaluation parameter to estimate the influence of the kinematic viscosity. Cavitation factors at transition points from continuous occurrence to intermittent occurrences were also evaluated as representative points where vortex cavitation occurs. Then, the occurrences of vortex cavitation were expressed as a relation between the cavitation factor at transition points and the non-dimensional circulation. As a result, it was clarified that the cavitation factor at transition points increased linearly in relatively small non-dimensional circulation, while it was nearly constant in relatively large non-dimensional circulation.

 Since the occurrence probability of multiple steam generator tube rupture (MSGTR) in a PWR is considered to be low, analytical or experimental investigation to prevent such accidents has not been performed explicitly. As new Japanese regulations require continuous effort to enhance the safety of nuclear power plants, low-probability but high-consequence events such as MSGTR should be taken into account to increase safety. In this study, a thermal-hydraulic analysis of multiple tube ruptures in a steam generator (SG) or all the SGs in a station blackout (SBO) situation was performed using the RETRAN-3D code in order to clarify the plant behavior during an MSGTR event and to contribute to risk reduction. The results show that a water supply function to SGs is important to cope with accidents involving MSGTR+SBO to prevent core damage. Furthermore, if the auxiliary feed water system loses its function when 10 tubes rupture in an SG, the time to core exposure is within 1 h and much shorter than that for SGTR, i.e., single tube rupture, in an SG. Therefore, in order to prevent core damage during MSGTR, it is desirable to have an alternative water injection system to quickly replace the auxiliary feed water system.