The existing maintenance program is focused on time-based maintenance to inspect and repair components according to maintenance period, rather than condition-based maintenance or predictive maintenance. The preventive maintenance template of the steam turbine has been developed for optimizing maintenance procedure and improving reliability and availability of the steam turbine of nuclear power plants based on EPRI PM template methodology and EPRI technical reports about preventive maintenance.

This paper describes the Visual Tool for automatic ultrasonic examination that is under developing as a part of the project for development of automatic ultrasonic wave acquisition and analysis program. This tool that is supported by various image processing techniques will be adopted to detect the flaws in the component and piping welds in NPP. Visual Tool will enhance the integrity of nuclear power plant. The object of this paper is to address the Visual Tool which is developing for automatic ultrasonic inspection of welds in NPP.

Cast austenitic stainless steel(CASS) is used in the primary cooling piping system of nuclear power plant for it's relative low cost, corrosion resistance and easy of welding. However, the coarse-grain structure of cast austenitic stainless steel can strongly affect the inspectability of ultrasonic testing. The major problems encountered during inspection are beam skewing, high attenuation and high background noise of CASS component. So far, the best inspection performance involving CASS components have been achieved using low frequency TRL(Transmitter/Receiver side-by-side L wave) angle beam probe. But TRL technique could not detect shallow defect and it contains an uncertainty for sizing capability. Currently, most of researchers are studying to overcome these challenge issue. In this study, low-frequency phased array TRL technique used to detect and sizing the flaws in CF8A cast austenitic stainless steel.As conclusion, we could detect and size not only axial flaw but also circumferential flaw using low frequency phased array technique.

To enhance the reliability of ultrasonic testing system, MOST(Ministry of Science and Technology) bulletin 2004-13 was published in 2004.6 requiring performance demonstration in ultrasonic examination and MEST(Ministry of Education, Science and Technology) 2009-37 was published in 2009.9(formerly MOST bulletin 2004-13). In order to meet the MEST bulletin requirement and increase the reliability of ultrasonic testing, all the ultrasonic examination procedure and equipment should be performance demonstrated and qualified. In this paper, generic procedure for the ultrasonic examination of dissimilar metal welds are described and new technique are introduced.

Ultrasonic inspection of heavy walled cast austenitic stainless steel(CASS)welds is very difficult due to complex and coarse grained structure of CASS material. The large size of anisotropic grain strongly affects the propagation of ultrasound by severe attenuation, change in velocity, and scattering of ultrasonic energy. therefore, the signal patterns originated from flaws can be difficult to distinguish from scattered signals. To improve detection and sizing capability of ID connected defect for heavy walled CASS piping welds, the low frequency segmented TRL Pulse Echo and Phased Array probe has been developed. The experimental studies have been performed using CASS pipe mock-up block containing artificial reflectors(ID connected EDM notch). The automatic pulse echo and phase array technique is applied the detection and the length sizing of the ID connected artificial reflectors and the results for detection and sizing has been compared respectively. The goal of this study is to assess a newly developed ultrasonic probe to improve the detection ability and the sizing of the crack in coarse-grained CASS components.

The stay cylinder weld at the steam generator of Korean Standard Nuclear Power Plants is safety class I component and is subjected to be inspected by the volumetric examination such as ultrasonic method. As accessibility of this area is limited due to the narrow space and high radiation, the existing manual inspection method involves various difficulties. Moreover operators may be exposed to internal contamination by contaminated dust during the surface buffing process to improve the inspection reliability of this area. Recently the new automatic inspection system for stay cylinder welds has been developed. The inspection system basically consists of a driving assembly, data acquisition device and signal processing units. The driving assembly is classified by 1) the scanner for inspecting and buffing the weld, 2) pillars for guiding the scanner and 3) the base frame for loading and supporting pillars. The scanner has 4 sensor modules to inspect in 4 refracted angles and 4 incident directions. These components can be inserted into the skirt of the stay cylinder through the manway hole and assembled easily by one-touch in the skirt. Data acquisition device and signal processing units developed in previous works are also newly upgraded for better processing of data analysis and evaluation. The system has been successfully demonstrated not only in the mock-up but also in the field. In this paper, newly developed inspection system for the stay cylinder weld of the steam generator is introduced and their field applications are discussed.

The piping failure probability of the reactor coolant system in Kori unit 1 was evaluated considering stress corrosion cracking. The P-PIE program (Probabilistic Piping Integrity Evaluation Program) developed in this study was used in the analysis. The effect of some variables such as oxygen concentration during start up and steady state operation, and operating temperature, which are related with stress corrosion cracking, on the piping failure probabilities was investigated. The effects of leak detection capability, the size of big leak, piping loops, and reactor types on the piping failure probability were also investigated. The results show that (1) LOCA (loss of coolant accident) probability of Kori unit 1 is extremely low, (2) leak probability is sensitive to oxygen concentration during steady state operation and operating temperature, while not sensitive to the oxygen concentration during start up, and (3) the piping thickness and operating temperature play important roles in the leak probabilities of the cold leg in 4 reactor types having same inner diameter.

Fatigue has been known as a major degradation mechanism of ASME class 1 components in nuclear power plants. Fatigue damage could be accelerated by combined interaction of several loads and environmental factors. However, the environmental effect is not explicitly addressed in the ASME S-N curve which is based on air at room temperature. Therefore many studies have been performed to understand the environmental effects on fatigue behavior of materials used in nuclear power plants. As a part of efforts, we performed low cycle fatigue tests under various environmental conditions and analyzed the environmental effects on the fatigue life of SA508 Gr. 1a low alloy steel by comparing with higuchi's model. Test results show that the fatigue life depends on water temperature, dissolved oxygen and strain rate. But strain rate over 0.4%/s has little effect on the fatigue life. To find the cause of different fatigue life with ANL's and higuchi's model, another test performed with different heat numbered and heat treated materials of SA508 Gr. 1a. On a metallurgical point of view, the material with bainite microstructure shows much longer fatigue life than that with ferrite/pearlite microstructure. And the characteristics of crack propagation as different microstructure seem to be the main cause of different fatigue life.

PHE(Process Heat Exchanger) is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR(Very High Temperature Reactor) to the chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute established the gas loop for the performance test of components, which are used in the VHTR, and they manufactured a PHE prototype to be tested in the loop. In this study, as part of the high-temperature structural-integrity evaluation of the PHE prototype, which is scheduled to be tested in the gas loop, we carried out high-temperature structural-analysis modeling, thermal analysis, and thermal expansion analysis of the PHE prototype. The results obtained in this study will be used to design the performance test setup for the PHE prototype.

The plugging rate of reheater tubes of Wolsung unit 1 nuclear power plant has been increased by corrosion and erosion since 1990. As the dimensions of the new first stage reheater bundle tubes which were supplied by Hanjung company to replace were different from old one, numerical calculations are carried out for flow and heat transfer in the reheater bundle tubes of the N.P.P. Numerical calculations consists of thermal performance, drain line pressure drop, flow change by pressure drop of line, stress analysis of finned tubes and analysis of flow induced vibration. Computational analysis using heat transfer research institute program is adopted to verify the results of the numerical calculations. It contains the evalution of performance in the system with view to location of the new reheater bundle and it shows the differences between the numerical calculation results and heat transfer research institute program output.

During the period of reinforcement work for the licensing renewal of CANDU NPP, the fuel channels, Calandria tubes and feeders of CANDU Reactor are replaced. The remote visual inspection of Calandria internal is also performed during the period of reinforcement work. This period is a unique opportunity to inspect the inside of the Calandria. The visual inspection for the Calandria vessel and its internals of Wolsong NPP Unit 1 was performed by Nuclear Engineering & Technology Institute(NETEC) of KHNP. To perform this inspection, NETEC developed equipment applied new technology such as the synchronization of 3D CAD, automatic alignment and control system. The inspection confirmed that the Calandria integrity of Wolsong NPP Unit 1 is perfect.

The feeder pipe of Main Primary Heat Transfer System in Wolsong Nuclear Power Plant was inspected by the Ultrasonic Phase Array technique in 2010. It is the first time to apply this method to the construction at Nuclear Power Plant in Korea. The time required for UT technique is less than RT method. The UT method doesn't need to evacuate personnel who works nearby inspecting area and doesn't need to wait developing of film. For these reasons, the UT method is the fastest method among the volumetric inspections. As a result of the examination, it became clear that main defect of the feeder pipe is the Lack of fusion in the welded area. Moreover, the rate of defect was reduced gradually as improvement of welder's skill. If welding machine has problem, the defect has tended to same pattern(occurred same position in the welding area) but these defects were founded without specific rules. For these reasons, the creation of defect is dependent on the skill of worker not on the automatic welding machine. This evaluation of defect signal and collecting data would be useful to further examination in ISI.

Domestic nuclear steam generators with Alloy 600 HTMA tubes have experienced axial cracking at eggcrate tube support plates(TSPs). The axial stress corrosion cracks were observed at the crevice between outside of tubes and eggcrate TSPs. The root cause of axial cracking was investigated by thermal hydraulic analysis and sludge distribution diagnosis. It is suggested that deposition of sludge at eggcrate TSPs could increase the outside surface temperature of tube and promote the enrichment of impurities at crevice, and thus accelerate cracking. Additionally strategy for reducing the sludge ingress to steam generators is discussed.