A failure or degradation of reactor upper head penetration is a troublesome problem at Nuclear Power Plants. A flaw in the reactor upper head penetration can result in unplanned plant shutdown for repair, and cause serious economic losses on the plants. Consequently, a detection of flaws is a matter of more importance. Until now, only the base metal, not including J-groove weld, in reactor upper head penetration has been inspected in accordance with 10 CFR 50.55a and ASME code case N-729-1 requirements. Accordingly, it is rather difficult to detect manufacturing defects and repair defects in J-groove weld. This paper presents a case study on the application of Time of Flight Diffraction UT technique to examine the J-groove weld in reactor head penetration using reactor head penetration mockup with artificial flaws. We expect that this study result will offer a way to understand the non-destructive examination technology for J-groove weld in reactor upper head penetration.

In this study, the performance test was carried out under various operation conditions targeting four representing types of lubricant which are mostly used in nuclear power plants, and mutually compared the characteristic and performance of lubricant. Especially, introducing the concept of the thread friction coefficient (hereinafter, TFC), which makes the friction relation between the stem nut and stem screw dimensionless. The test was performed to compare the lubricant performance for the four kinds of lubricant (Texaco, Alvania, Mobilux, MOVLL). In a test of the room temperature stem, the TFC of MOV Long Life shows the lowest value, next to Alvania EP2, Texaco EP2, and Mobilux EP0 in that order. And in a test of the high temperature stem, the TFC of Texaco EP2 shows the lowest, next to MOV Long Life, Alvania EP2, and Mobilux EP0 in that order. From the test result of the aging condition, three types of lubricant (MOV LL, Texaco EP2, Alvania EP2) show similar patterns up to 36 months, but in 60 months, the TFC of lubricant are increased rapidly.

Steam generator is one of the most important component of nuclear power plant, and it's integrity and reliability are to be assured to high level by pre-service inspection and in-service inspection. To improve the reliability of steam generator heat exchanger tubes and to secure the management of nuclear power plant safely, KHNP CRI recently has developed eddy current testing system for steam generator. KHNP CRI have performed a series of experimental verification and field application to confirm the performance of the developed ECT system in accordance with ASME Code requirements. The ECT system consists of a remote data acquisition unit, an ECT signal acquisition and analysis software, a water chamber robot controller and a probe push-puller. In this paper, we will details of the developed ECT system and the software and their experimental performance. And also we will report the field applying performance and the issues for further steps.

During refueling Outage, In-service inspections(ISIs) for the Nuclear Power Plant components are mandatory requirement in accordance with ASME Code Sec. XI. Especially, in current ultrasonic testing is one of the most important NDT techniques that are used for volumetric examination methods for nuclear power plant components, and accurate sizing of flaw indication by UT is essential to assure the integrity of the components. However, ASME code specifies minimum requirement for vessel examination procedure, and so far many different flaw sizing approaches have been tried to apply. Through the Round Robin Test(RRT), the accuracy of ultrasonic flaw sizing using DAC techniques was measured with the mock-ups simulating typical pressure vessel welds. These mock-ups contain artificially introduced flaws of known size and location. This paper shows experimental comparison data on the accuracy of techniques using such as 6dB drop, 50%DAC, 20%DAC and 20%DAC with beam spread correction, and also shows that diverse DAC techniques can be effectively applied to the assessment of the flaw sizing for pressure vessel welds in the stage of welding and fabrication.

In the present study, complex leakage probabilities of nuclear pipes due to fatigue and stress corrosion cracking are evaluated by using the PINTIN(Piping INTegrity INner flaws) that is developed based on the existing PRAISE(Piping Reliability Analysis Including Seismic Events) program. With regard to the aging and crack instability, small leak and big leak probabilities are calculated for several pipes in a reactor coolant system of domestic nuclear plant. Moreover, sensitivity analysis is also performed to find out the effect of parameters for the leakage of pipes, which shows the coolant temperature is the most influencing parameter.

Pipe wall-thinning by flow-accelerated corrosion (FAC) and various types of erosion is a significant and costly damage phenomenon in secondary piping systems of nuclear power plants (NPPs). Most NPPs have management programs to ensure pipe integrity due to wall-thinning that includes periodic measurements for pipe wall thicknesses using ultrasonic tests (UTs). Nevertheless, thinning evaluations are not easy because the amount of thickness reduction being measured is often quite small compared to the accuracy of the inspection technique. U.S. Electric Power Research Institute (EPRI) had proposed Total Point Method (TPM) as a thinning occurrence evaluation method, which is a very useful method for detecting locally thinned pipes or fittings. However, evaluation engineers have to discern manually the measurement data because there are no numerical algorithm for TPM. In this study, numerical algorithms were developed based on non-parametric and parametric statistical method.

The fatigue behavior of thermal butt fusion in safety class III high-density polyethylene (HDPE) buried piping for nuclear power plants was investigated using load-controlled bending fatigue on four-point bend test specimens. Based on the results, the presence of thermal butt fusion beads was confirmed to reduce the fatigue lifetime in the low- and medium-cycle fatigue regions while having a negligible effect in the high-cycle fatigue regions.

In many cases, the field fracture mechanism of the thermoplastic pipe is considered as either brittle or environmental fractures. Thus the estimation of the lifetime by modeling slow crack growth considering such fracture mechanisms is required. In comparison of the some conventional and empirical equations to explain the slow crack growth rate such as the Paris' law, the crack layer theory can be used to simulate the crack and process zone growth behaviors precisely, so the lifetime of thermoplastic pipe can also be accurately estimated. In this study, the modified crack layer theory for the stress corrosion cracking (SCC) of high density polyethylene is introduced with detailed algorithm. The oxidation induction time of the HDPE is also considered for the reduction of specific fracture energy during exposed to chemical environments. Furthermore, the parametric study for an important SCC parameter is conducted to understand the slow crack growth behavior of SCC.

Local wall thinning and integrity degradation caused by several mechanisms, such as flow accelerated corrosion, cavitation, flashing and/or liquid droplet impingement, is a main concern in secondary steam cycle piping system of nuclear power plants in terms of safety and operability. Thinned pipe management program (TPMP) has being developed and optimized to reduce the possibility of unplanned shutdown and/or power reduction due to pipe failure caused by wall thinning. In this paper, newest technologies, standards and regulations related to the integrity assessment, repair and replacement of thinned pipe component are reviewed. And technical improvement items in TPMP to secure the reliability and effectiveness are also presented.

3D FEM modeling based on 3D CAD data has been performed to evaluate the efficiency of CP system in a real operating nuclear power plant. The results of it successfully produced sophisticated profiles of electrolytic potential and current distributions in the soil of an interested area. This technology is expected to be a breakthrough for detection technology of damages on buried pipes when it comes into combining with a brand of area potential earth current (APEC) and ground penetrated radar (GPR) technologies. 2D current distribution and 2D current vectors on the earth surface from the APEC survey will be used as boundary conditions with exact 3D geometry data resulting in visualization of locations and extents of corrosion damages on the buried pipes in nuclear power plants.

Time domain response analysis for vibro-impact nonlinear behavior of multi-span tube with loose supports was performed using commercial FEA code and user subroutine. Support geometry of multi-span tube with a finite gap is realistically modeled by analytical rigid surface. Model of hydrodynamic force is based on the Qusai-steady model which accounts for the inclined angle of relative flow velocity and time delay between flow force and resulting tube motion. During tube vibration from flow loading, impact and friction at the support location is simulated using commercial FEA code with master slave contact algorithm. Analysis results has reasonable agreement with those of references and test experience. Plan of further refinement of analysis model and future test verification is briefly introduced.