Seal welding is a repair method when crack by Stress Corrosion Cracking (SCC) is found in weld rtion of core internals including pressure b oundary component in nuclear power plant. Seal lding is welded directly on surface of existing SCC crack and prevents SCC propagation by la ting SCC crack from reactor water environment. In this study, development of seal welding hnology using automatic TIG Welding was executed, and the applicability of seal welding method CRD stub-tube weld portion was confirmed by ex...
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Development of Seal Welding Repair Technology for CRD Stub-tube

In a previous paper, the authors discussed maintenance management suitable to R&D-level nuclear power plants, and made proposals. One of key proposals was to enhance knowledge on potential degradation mechanisms specific to a new reactor type by accumulating operational experience. It was noted that validation of R&D knowledge in actual plant conditions is important. In this study, a method for identifying degradation mechanisms to be validated by the PIRT (Phenomena Identification and Ranking Table) p...
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Study on a Method for Identifying Degradation Mechanisms to be Validated in Maintenance Management of R&D-Level Nuclear Power Plants Using the PIRT Process

A design study of an advanced sodium-cooled fast reactor (Advanced-SFR) has been conducted in Japan Atomic Energy Agency (JAEA). Hot sodium from the fuel assembly can mix with cold sodium from the control rod (CR) channel and the blanket assemblies at the bottom plate of the Upper Internal Structure (UIS). Temperature fluctuation due to mixing of the fluids at different temperature between the core outlet and cold channel may cause high cycle thermal fatigue on the structure around the bottom of UIS. A wate...
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Water Experiments on Thermal Striping Phenomena at the Core Outlet of an Advanced Sodium-cooled Fast Reactor? (1) Proposal of Countermeasures to Mitigate Temperature Fluctuations around Control Rods?

Focusing on the thermal striping phenomena that occurs at a bottom of the internal structure of an advanced sodium-cooled fast reactor (Advanced-SFR) that has been designed by the Japan Atomic Energy Agency, a water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of Upper Internal Structure (UIS). In the previous paper, we reported the effe...
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Water Experiments on Thermal Striping Phenomena at the Core Outlet of?an Advanced Sodium-cooled Fast Reactor? (2) Proposal of Countermeasures to Mitigate Temperature Fluctuations?around Radial Blanket Fuel Assemblies?

When a crack is detected in nuclear reactor core internals, it is important to survey details of a crack. In this study, to survey nuclear reactor core in ternals simply, new electrolytic etching method that applies “Gel Electrode” and underwater microscope were developed. To apply “Gel Electrode”, we can see fusion boundary and grain boundary after electrolytic etching. To use underwater microscope, we can see surface figures of nuclear reactor core internals directly without taking boat samples or repli...
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Development of Underwater Remote Inspection Technique for Nuclear Reactor Core Internals

In a process of hot and cold fluid mixing, temperature fluctuation in the fluid can occur. When this fluctuation is caused by unstable fluid mixing process, the surrounding structure would be damaged by high-cycle thermal fatigue. This phenomenon is observed in various nuclear power plants, which means it is important to investigate the relation between the fluid mixing process and the thermal fatigue. When there exists a 90-degree bend i n the upstream of the mixing area, the fluid mixing process becomes ...
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Estimation on the thermal hydraulic characteristics in a non-isothermal fluid mixing