Hiroshi Hirayama

Outline of the seismic buckling design guideline of an FBR — a tentative draft

Abstract Central Research Institute of Electric Power Industry (Japan), commissioned by the Ministry of International Trade and Industry, is carrying out the Demonstration Test and Research Program of Buckling of FBR (FY 1987-FY 1993). The first half of the research program was finished after establishing a seismic buckling design guideline (a tentative draft). The purpose of this paper is to describe the dynamic buckling characteristics of FBR main vessels and the outline of the rationalized buckling design guideline for seismic loadings.

Experimental Study on Thermal Striping in Mixing Tees With Hot and Cold Water

Thermal striping tests in mixing tees with hot and cold water were conducted for three types of flow conjunctions in order to establish an evaluation method for high-cycle thermal fatigue of piping systems. Two kinds of examinations were planned. The preliminary tests were flow visualization tests carried out using acrylic pipes to obtain flow pattern characteristics and flow temperature fluctuations. The main tests were temperature fluctuation measurement tests carried out using metal pipes to evaluate the unsteady heat transfer coefficient based on measured temperature fluctuations of fluid and pipe wall. This paper reports visualization test results. The flow patterns were visualized by injection of methylene blue and compared with flow analysis results by the k-e turbulence model. Temperature fluctuations of fluid 3mm from the inner pipe wall were measured with C-A thermocouples. Fundamental features such as locations with a large fluctuating temperature, the fluctuating temperature amplitude and its frequency were identified.Copyright

Development of elevated temperature structural design guide for Japanese demonstration breeder reactor

Abstract Programs to develop the “elevated temperature structural design guide for the demonstration fast breeder reactor” (DDS) in Japan have been conducted since 1987. The DDS is to be developed on the basis of the “elevated temperature structural design guide for class 1 components of prototype fast breeder reactors” (ETSDG), by considering structural and material features of the demonstration fast breeder reactor (DFBR) and incorporating results of the latest R&D. This paper describes the progress of the R& D concept of the DDS, and discusses some typical results of current studies on the DDS.

Proposal of a new estimation method of thermal ratchetting behavior of fast breeder reactor components

Abstract A new method for estimating the thermal ratchetting behavior of a cylinder subjected to an axially moving temperature distribution is proposed in this paper. This method considers the contribution of both the hoop membrane stress and the axial bending stress to the ratchetting behavior. Work hardening of the material also is considered for the stress-strain behavior which is assumed in the estimation method. Results predicted by this method agree well with the results obtained by finite element modelling and experimental results.

Crack Propagation Analysis Procedure Using FEM Applied to the Three-Dimensional Stress Field

Recently, several cracks have been found on the weld joints of Boiling Water Reactor (BWR) core shrouds during inspection. In order to ensure the continuous operation of nuclear power plants, it is necessary to assess the structural integrity of core shrouds with cracks on the weld joints. In general, a crack propagates in a complicated manner according to three-dimensional stress field and it is difficult to predict crack propagation direction and crack shape change. Usually, half ellipsoid crack shape is assumed and the propagation of the crack is calculated in the constant direction for assessment. In this study, crack propagation analysis procedure using the Finite Element Method (FEM) is developed for general shaped crack, and the procedure is verified by experiments. In this procedure, it is assumed that the crack propagates according to the maximum J-integral under three-dimensional stress fields and the re-mesh technique is used in the FEM analysis in order to calculate crack shape variation during propagation. The fatigue crack propagation tests under cyclic tensile load were performed to verify the analysis procedure. The specimens are made of a plate from 316SS and designed to generate non-uniform stress distribution on the crack front in order to observe continuous crack propagation direction change.Copyright

A study of the structural integrity of the core support structure of a fast breeder reactor

This paper reports on the core support structure of a fast breeder reactor supports the fuel assemblies, supplies sodium coolant to the fuel assemblies, and maintains the insertability of control rods even during an earthquake. The core support structure is designed as a box fabricated of welded plates, ribs, and cylinders that distribute the load in a diverse manner, in order to reduce the weight and to fulfill safety-related functions. This box structure was not adopted in the Monju prototype reactor; thus, a method to evaluate the structural integrity of this structure must be developed. To prepare design guidelines, structural integrity was studied in accordance with the requirements and features of the box structure. From the results of these experiments, the crack growth rate was evaluated and incorporated into the structural integrity evaluation method. Finally, the structural integrity of the core support structure of the Japanese demonstration reactor is evaluated by this method.