This paper describes a newly-developed analysis methodology in comprehensive vibration assessment program (CVAP) of reactor internals to develop a valid-prototype for the design of nuclear power plants. The new analysis methodology developed in this study will be confirmed through a scale model testing (SMT). Based on the measurements obtained from dynamic pressure transducers in the SMT, a new non-dimensional equation is developed to apply the forcing functions at reactor internals for the prototype. In addition to the new non-dimensional equation, a computational fluid dynamics(CFD) is used to develop the application of the hydraulic loads at reactor internals for the prototype.

Only about 10% of selected equipment in nuclear power plants are monitored by wiring to address failures or problems caused by vibration. The purpose is primarily for preventive maintenance, not for predictive maintenance. This paper shows that vibration monitoring and diagnosis using Industrial 4.0 enables the complete predictive maintenance for all vibrating equipments in nuclear power plants with the convergence of internet of things; wireless technology, big data through periodic collection and artificial intelligence. Predictive maintenance using wireless technology is possible in all areas of nuclear power plants and in all systems, but it should satisfy regulatory guides on electromagnetic interference and cyber security.

Additive manufacturing technology, called as 3D printing, is one of fourth industrial revolution technologies that can drive innovation in the manufacturing process, and thus should be applied to nuclear industry for various purposes according to the manufacturing trend change in the future. In this paper, we performed tensile tests of 3D printed stainless steel 316L as-built specimens manufactured by two types of technology; DED (Directed Energy Deposition) and PBF (Powder Bed Fusion). Their mechanical properties (tensile strength, yield strength, elongation and reduction of area) were compared. As a result of comparison, the mechanical properties of the PBF specimens were slightly better than those of DED specimens. In the same additive type of specimens, the tensile and yield strength of specimens in the X and Y direction were higher than those in the Z direction, but the elongation and ROA were lower.

The defected baffle former bolts of Kori unit 1 were withdrawn to analyze the cause of damage and gamma-ray measurement is being scheduled. Prior to that, in order to calculate the specific radioactivity value of the baffle former bolt, a radioactivity calculation method considering the actual operation history of the nuclear power plant is introduced and the calculation results are shown. In particular, the radioactivity calculation method considering the operation history is obtained by defining the monthly contribution factor from the actual monthly operation history. As a result, the results considering operation history are 16-28% lower than the general radioactivity calculation results. These results can contribute to establish a reasonable but economical strategy when planning nuclear power plant decommissioning.

Pressure tubes are the main components of PHWR core and serve as the pressure boundary of the primary heat transport system. However, because pressure tubes have changed their geometrical dimensions under the severe operating conditions of high temperature, high pressure and neutron irradiation according to the increase of operation time, all dimensional changes should be predicted to ensure that dimensions remain within the allowable design ranges during the operation. Among the deformations, the diameter expansion due to creep leads to the increase of bypass flow which may not contribute to the fuel cooling, the decrease of critical channel power and finally the deration of the power to maintain the operational safety margin. This study is focused on the modeling of the expansion of the pressure tube diameter based on the operating conditions and measured diameter data. The pressure tube diameter expansion was modeled using the neutron flux and temperature distributions of each fuel channel and each fuel bundle as well as the measured diameter data. Although the basic concept of the current modeling approach is simple, the diameter prediction results using the developed methodology showed very good agreement with the real data, compared to the existing methodology.

Lots of researches have been conducted on in-vessel retention (IVR) to prevent or mitigate severe accident in nuclear power plants. Various methodologies were proposed and the external reactor vessel cooling was selected as a part of promising IVR strategy. In this study, the strategy is strengthened by enhancing the natural circulation performance through the adoption of insulation in the reactor cavity. A thermal analysis was carried out based on an assumed accident scenario and its results were used as boundary conditions for subsequent seven flow analysis cases. By comparing the natural circulation performance, effects of annular gaps and insulation shapes on the mass flow rate and flow velocity were quantified. The improvement in cooling performance can be reflected in actual design via detailed assessment.

The RCP is one of the main components in nuclear power plants and plays an important role in circulating coolant to the RCS system. Currently, nuclear plants are monitored using various monitoring systems. However, since they operate independently according to their functional purpose, it is not able to analyze vibration and operation/performance information comprehensively, and thus failure diagnosis accuracy is limited. In addition, these systems do not provide some important information (such as fault type, parts and cause) necessary for emergency actions, but provide only alarm information. To improve these technical problems, this study proposes a diagnosis technique (M/L, Rule-based model, Data-driven model, Narrow band model) and methodology for comprehensive analysis.

The welding part of the valve needs immediate action when leakage occurs due to cracks or damage. In order to repair leakage of the welding part, the valve must be separated from the pipe or replaced with a new valve. However, it is difficult to remove the valve while operating the power plant. This study presents a method to remove leakage by precisely processing the gap between the clamp and the incision part within 0.1mm while installed in the pipe system. If the external leakage is removed using a clamp on the welding part without removing the valve during operation, the time and cost required for maintenance can be reduced.