In this paper, the regulations and technical standards related to qualification of safety grade equipment in nuclear power plants are critically reviewed with the qualification procedure in terms of structures, systems, and equipment in nuclear power plants. These facilities should be designed and constructed to protect from natural conditions or disasters and to perform their safety functions even in case of postulated accidents. Equipment Qualification is to demonstrate that the safety related equipment is designed and constructed to perform their safety functions under normal and accident conditions. It is classified into environmental qualification and seismic qualification.

Perforated plates are used for the steam generator tube-sheet and the Reactor Vessel Closure Head in the Nuclear Power Plant. The ASME code, Section III Appendix A-8000, addresses the analysis of perforated plates, however, this analysis is only limited to the flat plate with a triangular perforation pattern. Based on the concept of the effective elastic constants, simulation of flat and spherical perforated plates and their equivalent solid plates were carried out using Finite Element Analysis (FEA). The isotropic material properties of the perforated plate were replaced with anisotropic material properties of the equivalent solid plate and subjected to the same loading conditions. The generated curves of effective elastic constants vs ligament efficiency for the flat perforated plate were in agreement with the design curve provided by ASME code. With this result, a plate with spherical curvature having perforations can be conveniently analyzed with equivalent elastic modulus and equivalent Poisson's ratio.

The Advanced Power Reactor 1400 (APR1400) Steam Generator (SG) uses alloy 690 as a tube material and SA-508 Grade 3 Class 1 as a tubesheet material to form tube-to-tubesheet joint through hydraulic expansion process. In this paper, the residual stresses in the SG tube-to-tubesheet contact area was investigated by applying Model-Based System Engineering (MBSE) methodology and the V-model. The use of MBSE transform system description into diagrams which clearly describe the logical interaction between functions hence minimizes the risk of ambiguity. A theoretical and Finite Element Methodology (FEM) was used to assess and compare the residual stresses in the tube-to-tubesheet contact area. Additionally, the axial strength of the tube to tubesheet joint based on the pull-out force against the contact joint force was evaluated and recommended optimum autofrettage pressure to minimize residual stresses in the transition zone given. A single U-tube hole and tubesheet with ligament thickness was taken as a single cylinder and plane strain condition was assumed. An iterative method was used in FEM simulation to find the limit autofrettage pressure at which pull-out force and contact force are of the same magnitude. The joint contact force was estimated to be 20 times more than the pull-out force and the limit autofrettage pressure was estimated to be 141.85MPa.

In nuclear power plant operation and spent fuel canisters, it is necessary to provide a sound technical basis for the safety and security of long-term operation and storage respectively. Recently, the peening technology is being discussed and the technology will be adopted to ASME Section III, Division 1, Subsection NX (2019 Edition). The peening is prohibited in current edition, but it will be approved in 2019 Edition and adopted. However, Surface stress improvement techniques such as the peening is used to mitigate SCC susceptible in operating nuclear plants. Although the peening will be approved to ASME CODE, there are no performance criteria listed in the 2019 edition. The Korean International Working Group (KIWG) formed a new Task Group named "Advanced Surface Stress Improved Technology". The task group will develop a CODE CASE to address PWSCC(Primary Water Stress Corrosion Cracking) and CISCC(Chloride Induced Stress Corrosion Cracking) for new ASME Section III components. TG-ASSIT was started to make peening performance criteria for ASME Section III (new fabrication) applications. The objective of TG-ASSIT is to gain consensus among the relevant Code groups that requirements/mitigation have been met.

An 1,500MW advanced power reactor required the standard design approval by a Korean regulatory body in 2014. The reactor has been designed to have a 4-train independent safety concept and a passive auxiliary feedwater system (PAFS). The full power risk or core damage frequency (CDF) of 1,500MW advanced power reactor has been reduced more than that of APR1400. However, the risk during the low power and shutdown (LPSD) operation should be reduced because CDF of LPSD is about 4.7 times higher than that of internal full power. The purpose of paper is to analysis design alternatives to reduce risk during the LPSD. This paper suggests design alternatives to reduce risk and presents sensitivity analysis results.

Unlike the welded pipes in the primary system of light water nuclear power plants being periodically inspected with in-Service inspection program, relatively small pipes with the outer diameter less than 2 inch have not been regularly inspected to date. However, after several failure reports on the occurrence of critical crack-like defects in small pipes, inspection for the small pipes has been more demanded because it could cause the provisional outage of nuclear power plants. Nevertheless, there's no particular method to examine the small pipes having access limitations for inspection due to various reasons; inaccessible area, excessive radiation exposure, hazardous surrounding, and etc. This study is to develop a reliable inspection technique using torsional and flexural modes of guided wave to detect defects that could occur in inaccessible area. The attribute of guided wave that can travel a long distance enables to inspect even isolated range of the pipe from accessible location. This paper presents a case study of the evaluation test on 3/4" small-bore pipes with guide wave method. The test result demonstrates the crack signal behavior and assures possibility to detect the crack signal in a flexural mode, which is clearly distinguishable from the symmetric structure signal in a torsional mode.

This work proposes a noncontact ultrasonic transducer for detecting wall-thinning of containment liner plates of nuclear power plants by measuring their thickness without physical contact. Because the containment liner plate is designed to prevent atmospheric leakage of radioactive substances under severe nuclear accident, its wall-thinning inspection is important for safety of nuclear power plants. Wall-thinning investigation of containment liner plates have been carried out by measuring their thickness with contact-type ultrasonic thickness gauge by inspectors and needs a lot of time and cost. As an alternative, an electromagnetic acoustic transducer measuring precisely thickness of containment liner plates without any physical contact or couplant was suggested in this research. A transducer generating and measuring shear ultrasonic waves in thickness direction was designed and wave field produced by the transducer was analyzed to verify the design. The working performance of the suggested transducer was tested with carbon steel plate specimens with various thicknesses. The test result shows that the proposed transducer can measure thickness of the specimens precisely without any couplant and implies that swift scanning of wall-thinning of containment liner plates will be possible with the proposed transducer.

The seismic capacity of electric cabinets in Nuclear Power Plants (NPPs) should be qualified before installation and be maintained during operation. However it can happen that identical devices cannnot be produced for replacement of devices mounted in electric cabinets. In case of when no In-Cabinet Response Spectrum (ICRS) is available for new devices, ICRS can be generated by using Finite Element Analysis (FEA). In this study we investigate structural response and ICRSs of battery charger which is supplied to NPPs. Test results on the battery charger are utilized in this study. The response is measured by accelerometers installed on the housing of the battery charger and local panels in the battery charger. Numerical analysis model is established based on resonant frequency search test results and validated by comparison with 2 types of earthquake testing results. ICRSs produced from the numerical model are compared with measured ICRSs in the seismic tests. Developed analysis model is a simple reduced model and anticipates ICRSs quite well as measured response in the tests overall despite of its structural limitation.

The APR1400 reactor may be operated for a long time under high temperature and pressure conditions, causing damage to the stud holes and causing stud bolts and holes to stick. The present practice is to manually remove the anti-sticking agent and foreign matter remaining in the APR1400 reactor stud hole and to visually check the surface condition of the thread to check the damage status of the threads. In the case of the APR1400 reactor stud holes, manually cleaning the threads increases the risk of radiation exposure and operator's fatigue. To avoid this, the autonomous mobile robot is used to automatically clean the reactor stud holes. The purpose of this study is to optimize the cleaning performance of the mobile robot by looking at the behavior of the surface roughness of the stud surface cleaned by the brush attached to the mobile robot due to changes in brush material, thickness of wire, and rotation speed. A microscopic approach to the surface roughness of the flank is needed to investigate the effects of the newly proposed brush of the autonomous mobile robot on the thread holes. According to this experiment, it is reasonable to use STS brush rather than Carbon one. Optimal operating conditions are derived and the safety of APR1400 reactor stud holes maintenance can be improved.

Actual stress distributions in the nozzle of a pressure vessel may not be in plane strain condition, implying that the crack-tip constraint condition may be relaxed in the nozzle. In this paper, a methodology for evaluating the fracture toughness of the ASME Code is presented considering the relaxation of the constraint effect in the nozzle of the reactor pressure vessel. The crack-tip constraint effect is quantified by the T-stress. The equation, which represent the relation between the fracture toughness in the lower constraint condition and the plane strain fracture toughness, is derived using the T-stress. This equation is similar to the method for evaluating the fracture toughness of the Master Curve for low constraint conditions. As a result of evaluating the fracture toughness considering the constraint effect in the reactor inlet, outlet and direct injection nozzles using the proposed equation, it was confirmed that the fracture toughness in the nozzles is higher than the plane strain fracture toughness. Applying the proposed evaluation methodology, it is possible to reflect the relaxation of the constraint effect in the nozzles of the reactor pressure vessel, therefore, the safe operation area on the pressure-temperature limit curve can be prevented from being excessively limited.