Hideyasu Furukawa

Post-test analysis of a 1:10-scale top slab model of ABWR/RCCV subjected to internal pressure

Abstract Construction of the first Advanced Boiling Water Reactor (ABWR) in Japan employing a reinforced concrete containment vessel (RCCV) was started in 1991. As RCCV itself is the first structure of its kind in Japan, thorough verification tests have been performed. This paper presents the results of simulation analysis of the Top Slab partial model of the RCCV subjected to internal pressure beyond design load. The Top Slab portion is complicated, being composed of a flat Top Slab, cylindrical wall and fuel pool girders, that its simulation analysis requires the evaluation of nonlinear structural behavior of reinforced concrete members due to membrane, bending and shear forces. This paper reports that Finite Element analysis with 3-D solid elements has given a good quantitative agreement between experimental and analysis results with respect to deformation, failure load and each nonlinear behavior.

Experimental study on RCCV of ABWR plant

Abstract This research relates to the application of a reinforced concrete containment vessel (RCCV) whose form can be freely selected to the development of an advanced BWR. The researchers on RCCVs are carried out recently, and good results are obtained. The RCCV employed in this research is different from conventional ones in structural elements and form. The purpose of this research is, therefore, to confirm the adequacy and reliability of the design method of the trial-designed RCCV, and to confirm the ultimate strengt and constructibleness. Basic, partial model and total model experiments were conducted. As a result of experimental and analytical studies, it was made sure that the trial-designed RCCV is safe and reliable at a design-load level. It was also confirmed that its ultimate strength fully ensures structural performance. Realizability was confirmed from consideration of the adequacy of the design method and constructibleness. This paper describes the results of the total model experiments planned for these researches.

Technical Report. Development of In-core Monitor Housing Replacement Technology.

Replacement technology of in-core monitor (ICM) housing of BWR was developed and applied in the operating Plant. This technology replaces stress corrosion cracking (SCC) susceptible existing ICM housing by new housing composed of SCC resistant material. Major procedure steps are removal of the existing ICM housing, welding and trimming of a new weld buildup, welding of a new ICM housing, and joint of a new ICM guide tube with the ICM housing using shape memory alloy couplings. Each procedure step is performed remotely with automated tools, which are installed from the reactor refueling floor, 26m down to the Reactor Pressure Vessel bottom, through core plate etc. After several years of development, this technology was applied to an ICM housing of Fukushima Daiich Unit #4 where a crack caused by SCC was found on October 1997. and site work was successfully completed in the end of January 1998.