Yuichi Koide

Full Scale Simulation-Based Study on Dynamic Response of BWR Fuel Assemblies Under Seismic Loading

The purpose of this study is to investigate dynamic response behaviors of fuel assemblies in a boiling water reactor (BWR) under seismic loading. The core of BWR consists of several hundreds of fuel assemblies. They are supported with both top guide and fuel support and are surrounded by coolant water. It is important to grasp their dynamic response behaviors under seismic loading for securing the structural integrity of the fuel assembly itself as well as for assessing control rod scrammability. In this study, we employ two different numerical simulation methods of acoustic fluid-structure interaction (AFSI) developed by the present authors independently. The one is a three-dimensional parallel finite element method for AFSI problems with solid elements based on a partitioned coupling approach, while the other is a finite element method of beam elements for fuel assemblies combining added mass matrix, which represents coupled inertia effects caused by coolant water.Both methods are first applied to a problem of 36 fuel assemblies for numerical verification, and then applied to a problem of 368 fuel assemblies for validation. The latter problem was set up based on the demonstration test performed by the NUPEC (Nuclear Power Engineering Corporation) in 1986. Both simulation results agreed well with each other in all cases, and the simulated results also agreed well with the experimental ones. In addition, we have precisely discussed seismic response behaviors of the fuel assemblies, which were not shown in the demonstration test. Accordingly, we conclude that the both developed simulation methods are powerful tools to grasp the precise behavior of fuel assemblies of BWR under seismic loading and to improve the seismic safety design of BWR core.Copyright

Modeling of Fuel Assemblies in a Boiling Water Reactor for Seismic Analysis

The purpose of this study is to develop a seismic analysis model of a group of fuel assemblies in a boiling water reactor and to confirm the validity of the developed model. Each fuel assembly was modeled as a beam on the basis of the finite element method. The mass matrix of the model includes an added mass matrix, which represents the coupled inertia effect caused by the coolant water, in order to simulate the coupled vibration of fuel assemblies. The added mass matrix was obtained by calculating the coefficient matrix of the acceleration vector and fluid force vector under the condition that each fuel assembly moves at unit acceleration. The validity of the model was confirmed by comparing the calculated results with experimental ones. The compared specimens for the experiments were full-scale mock-ups. The vibration characteristics of fuel assemblies in each case of 4 bodies and 368 bodies were compared. As a result of the comparison, the calculations of the frequency response were in agreement with the experimental results. Particularly, the calculation results on the resonance frequency were in good agreement, with an error of less than 2 percent, with the experimental ones. Furthermore, the calculated vibration characteristics of 368 fuel assemblies in the case of an earthquake, such as the excited vibration mode and phase characteristics, were in agreement with the experimental ones. We concluded that the developed model of fuel assemblies was applicable to seismic analysis of a boiling water core.Copyright

Dynamic Modeling of a BWR Control Rod Insertion System for Seismic Analysis and its Experimental Validation

The assessment of the seismic scrammability, which means the control rod insertability during a seismic event, is one of the most important design tasks for ensuring the seismic safety of nuclear power plants in Japan. This paper discusses the dynamic modeling of the control rod insertion behavior of a boiling water reactor (BWR) during an earthquake. A dynamic model of a control rod insertion system for BWR was developed based on multi-body dynamics. The coupled vibration behavior of the fuel assemblies in the fluid was modeled as an inertial coupling system. The effect of the interaction force between the control rod and the fuel assemblies was considered in a three-dimensional contact analysis. The hydraulic control unit and the control rod drive, which provide the control rod with drive force, were modeled in the concentrated parameter system. The model parameters, such as the friction coefficient between the control rod and the fuel assembly and the discharge coefficient of the scram piping, were obtained by conducting experiments. The validity of the model was confirmed by comparing the analytical results with the experimental ones. First, the validity of the fuel assembly model was verified through a comparison with the vibration testing in an underwater condition. It was confirmed that the calculation results for the frequency response of the fuel assembly were in good agreement with the experimental ones. Second, the validity of the modeling method of the drive system consisting of the hydraulic control unit and the control rod drive was verified through a comparison with the scram testing under non-vibration condition. The calculation results for the time history of the control rod insertion, the accumulator pressure, and the flow through the scram piping were in good agreement with the experimental ones. Finally, the validity of the modeling method of the whole system consisting of the fuel assemblies, the control rod, and the drive system was verified through a comparison with the scram testing under vibration condition. The calculation results for the time history of the control rod insertion stroke and the time delay of the insertion motion during an earthquake were in good agreement with the experimental ones. The results of these comparisons show that the developed analysis model can simulate the control rod insertion behavior during an earthquake.Copyright

Development of Evaluation Method to Determine Fragility of Power Plant Equipment Based on Bayesian Analysis

In the seismic design of nuclear power plant, it is recently considered to apply the probabilistic method in addition to the deterministic method. The deterministic method secures the seismic safety of a nuclear power plant by veryfying that the calculated stress on the equipment is lower than allowable stress in the design criteria. The seismic probability safety assessment (Seismic PSA) method secures seismic safety by veryfying that the probability of reactor core failure is lower than a certain value. In seismic PSA, it is important to estimate the fragility of the equipment for precise assessment. Therefore, in this paper we dealt with a method using the Bayesian approach to generate a fragility curve of power plant from a multi-specimen test, and investigated the influence of the sample size driving the fragility evaluation test