Hideki Yuya

Application of Stress Intensity Factors for Deep Surface Cracks to Crack Growth Evaluation

Flaw evaluation for nuclear power plants is conducted on the basis of a fitness-for-service code. For instance, the ASME Boiler and Pressure Vessel Code Section XI (ASME Section XI) and the JSME Rules on Fitness-for-Service for Nuclear Power Plants (JSME Code) prescribe a flaw evaluation procedure. In flaw evaluation, an aspect ratio of a detected surface crack is defined by a/l, where a is the crack depth and l is the crack length, and the aspect ratio a/l does not exceed 0.5. Therefore, a deep surface crack, which has an aspect ratio a/l greater than 0.5, is characterized as a semicircle with l = 2a. Meanwhile, deep surface cracks caused by stress corrosion cracking (SCC) have been detected in the Ni based alloy weld metal. Since the limit of the flaw characterization rule which is an aspect ratio a/l ≤ 0.5 seems to conduct to a conservative evaluation result for a deep surface crack, more rational and applicable flaw evaluation is required in order to eliminate surplus conservatism.In this study, a flaw evaluation procedure based on ASME Section XI or JSME Code is extended to deal with a deep surface crack. To evaluate crack growth behavior for a deep surface crack, coefficients to calculate stress intensity factors were evaluated by finite element analysis (FEA) and shown in tabular form on the basis of equations prescribed in ASME Section XI and JSME Code. To verify the applicability of proposed coefficients to crack growth evaluation, SCC crack growth behavior for a deep initial crack was evaluated by coefficients applied to the ASME Section XI procedure and a detailed FEA method. Applicability of coefficients to crack growth evaluation was verified through comparisons of crack growth behaviors for deep surface crack under various stress fields.Copyright

Feasibility Study on a Simple Method of Retrospective Neutron Dosimetry for Reactor Internals and Reactor Vessel

Nondestructive and simple retrospective dosimetry method was developed. Mercuric iodide or cadmium zinc telluride detector was used as a detector to measure 54Mn, 58Co, and 60Co that were created in reactor internals and vessels. Detector is surrounded by tungsten collimator which shield background gamma rays and detect gamma rays from desired measuring point. Photopeak count rate of this detector were calculated using the MCNP-4C code. This response was calculated for each reactor internal geometry that was irradiated by unit neutron flux which energy spectrum was calculated by the TORT code. Neutron fluence is calculated using precalculated response, measuring time, decay time and reactor power history. We can also get Displacement per Atom (DPA), hydrogen and helium production simultaneously using neutron spectrum precalculated in reactor vessel. Because the time and cost of this method are much less than the sampling method, we can measure at many positions and even get a contour map.

Studies on Flow and Pipe Wall Thinning and the Reduction in the Downstream of Orifice Nozzle

In a piping system of power plant, pipe wall thinning by Flow Accelerated Corrosion, FAC, Liquid Droplet Impingement Erosion, LDI, and Cavitation Erosion, C/E, are very serious problems because they give a damage and lead to the destructtion of the piping system[1]–[6] . In this study, the pipe wall thinning by FAC in the downstream of orifice nozzle, flow meter, is examined. Namely, the characteristics of FAC, generation mechanism, and prediction of the thinning and the reduction are made clear by experimental analysis. As a results, it was made clear that (1) the thinning is occurred mainly according to the size of the pressure fluctuation p′ on the pipe wall and the thinning can be estimated by it, and (2) the suppression of p′ can be realized by replacing the orifice to a taper shaped one having an angle to the upstream.Copyright