Crud is a corrosion deposit that forms in equipments and piping of nuclear reactor's primary systems. When crud circulates through the reactor's primary system coolant and adheres to the surface of the nuclear fuel cladding tube, it can lead to the Axial Offset Anomaly (AOA) phenomenon. This occurrence is known to potentially reduce the output of a nuclear power plant or to necessitate an early shutdown. Consequently, worldwide nuclear power plants have employed ultrasonic cleaning methods since 2000 to mitigate crud deposition, ensuring stable operation and economic efficiency. This paper details the system configuration of ultrasonic nuclear fuel cleaning equipment, outlining the function of each component. The objective is to contribute to the local domestic production of ultrasonic nuclear fuel cleaning equipment. Additionally, the paper introduces a method for accurately measuring the weight of removed crud, a crucial factor in assessing cleaning effectiveness and providing input data for the BOA code used in core safety evaluations. Accurate measurement of highly radioactive filters containing crud is essential, and weighing them underwater is a common practice. However, the buoyancy effect during underwater weighing may lead to an overestimation of the collected crud's weight. To address this issue, the paper proposes a formula correcting for buoyancy errors, enhancing measurement accuracy. This improved weight measurement method, accounting for buoyancy effects in water, is expected to facilitate the quantitative assessment of filter weights generated during chemical decontamination and system operations in nuclear power plants.

Environmental qualification is required for safety related electrical equipment under harsh environments located in nuclear power plants according to 10 CFR 50.49 and RG 1.89. As analog technology has recently been replaced by digital technology, NRC established RG 1.209 as a regulatory guideline for environmental qualification of safety related computer-based I&C system located in mild environments, requiring evaluation for electromagnetic compatibility, smoke exposure and type test for actual service conditions such as temperature and humidity. In this paper, the trend of environmental qualification for digital equipment is analyzed by comparing the environmental qualification requirements between digital and analog equipment.

Denting of steam generator (SG) tube is defined as the reduction in tube diameter due to the stresses exerted by the corrosion products formed on the outer diameter surface. This phenomenon is mostly observed in the crevices between SG tube and the top-of tubesheet or tube support plate. Despite the replacement of SG tube with Alloy 690, which has better corrosion resistance than Alloy 600, the denting of SG tube still remains a potential problem that could decrease the SG integrity. Deformation of SG tube by denting phenomenon can affect the surface properties and microstructure of SG tube. In this study, the effects of plastic deformation on surface properties and microstructure of Alloy 690 thermally treated (TT) tube was investigated by using the various analysis techniques. The plastic deformation of Alloy 690 increased the surface roughness and area. Many surface defects such as ripped surface and micro-cracks were observed on the deformed Alloy 690TT specimen. Based on the electron backscatter diffraction analysis, the dislocation density of deformed SG tube increased compared to non-deformed SG tube. In addition, the effects of changes in surface properties and microstructure of SG tube on general corrosion behavior were discussed.

The core shroud of rector vessel internals (RVI) of OPR1000 and ARP1400 is made of Type 304 stainless steel (SS) by bending and welding process that may induce high deformation and residual stress in manufacturing. This work aims to evaluate the susceptibility of stress corrosion crack (SCC) initiation of bent parts of RVI in high temperature primary water environment. For SCC initiation test, tensile specimens were fabricated from the 90 degree bent plate of Type 304 SS (DT specimen), that is an archived part of a Korean APR1400. After the SCC initiation test, the specimen surface was thoroughly examined by optical and scanning electron microscopy, and compared to the specimen fabricated from the as-received plate of Type 304 SS (AR specimen). The surface observation revealed that SCC initiated on the AR specimen surface in typical intergranular (IG) mode, while SCC on the DT specimen occurred in transgrannular mode as well as IG mode. It was also found that the size and number of SCC on the DT specimen were larger than that on the AR specimen. This was attributable to a strain-hardening during the bending process. To compare the susceptibility of SCC initiation, total crack density (TCD) was calculated from the total crack length divided by the measured area of AR and DT specimens. TCD of DT specimen was 4.6 times higher than AR specimen in average, indicating that higher possibility of degradation of bent parts of RVI for a long-term operation.

The best estimate plus uncertainty methodologies for loss-of-coolant accident analyses make use of the best-estimate codes and relevant experimental databases. Inherently, best-estimate codes have various uncertainties in the model parameters, which can be quantified by the dedicated experimental database. Therefore, this study was devoted to establishing procedures for identifying the input parameters of predictive models and quantifying their uncertainty ranges. The rod bundle heat transfer experiments were employed as a representative reflood separate effect test, and the TRACE code was utilized as a best-estimate code. In accordance with the present procedure for uncertainty quantification, the integrated list of the influential input parameters and their uncertainty ranges was obtained through local sensitivity calculations and screening criteria. The validity of the procedure was confirmed by applying it to uncertainty analyses, which checks whether the measured data are within computed ranges of the variables of interest. The uncertainty quantification procedure proposed in this study is anticipated to provide comprehensive guidance for the conduct of uncertainty analyses.

Research reactors are operated under ambient temperature and atmospheric pressure, which is much less severe conditions compared to those in typical nuclear power plants. Due to the high temperature, heat resistant materials such as austenite stainless steel should be used for the reactors in typical nuclear power plants. Whereas, as the effect of temperature is low for research reactors, materials with high resistance to neutron irradiation, such as zircaloy and beryllium, are used. Therefore, these conditions should be considered when performing integrity assessment for research reactors. In this study, a computational technique through finite element (FE) analysis was developed considering the operating conditions and materials of research reactor when conducting integrity assessment. Neutron irradiation analysis techniques using thermal expansion analysis were proposed to consider neutron irradiation growth and swelling in zirconium alloys and beryllium. A user subroutine program that can calculate the strain rate induced by neutron irradiation creep was developed for use in the commercial analysis program Abaqus. To validate the proposed technique and the user subroutine, FE analysis results were compared with hand-calculation results, and showed good agreement. Consequently, developed technique and user subroutine are suitable for evaluating structural integrity of research reactors.

Stainless 630 (or 17-4PH) is a precipitation-hardening martensitic stainless steel that has excellent mechanical properties and corrosion resistance. These characteristics make the STS630 to be used as a consisting material for various components such as spider, pin, spring, and spring retainer, of the control rod drive mechanism (CRDM) in pressurized water reactors (PWRs). In general, it is well known that the oxide layer of stainless steel consists of a duplex layer, a compact inner layer of FeCr₂O₄ spinel, and a coarse-grained outer layer of Fe₃O₄ spinel in PWR primary coolant condition. However, the characteristics of the oxide layer can be sensitively influenced by various water chemistry conditions such as temperature, dissolved oxygen, dissolved hydrogen, pH, pH adjuster type, and exposure time. In this work, we investigate the corrosion properties of the STS630 as a function of coolant temperature in an NH3 alkaline solution for its boron-free application in a small modular reactor, to confirm the feasibility for usage as a boron-free SMR structural material. As a result, oxide layer of corroded STS630 is consist of double-layer oxides consisting of a Cr-rich dense inner oxide and a Fe-rich polyhedral outer particles like as that in commercial PWR primary coolant. The corrosion rate of STS630 increases with increase in test time and temperature and the corrosion rate-time model equation was developed based on experimental data. Overall, it is expected that the results in this study provides useful data for the corrosion behavior of STS630 in alkaline environments, contributing to the development of selecting suitable materials for SMRs.

Cast austenitic stainless steels (CASS) and austenitic stainless steel weldments with a ferrite-austenite duplex structure are widely used in nuclear power plants, incorporating ferrite phase to enhance strength, stress relief, and corrosion resistance. Thermal aging at 290-325℃ can induce embrittlement, primarily due to spinodal decomposition and G-phase precipitation in the ferrite phase. This study evaluates the effects of thermal aging by collecting and analyzing various mechanical properties, such as Charpy impact energy, ferrite microhardness, and tensile strength, from various literature sources. Different model expressions, including hyperbolic tangent and phase transformation equations, are applied to calculate activation energy (Q) of room-temperature impact energies, and the results are compared. Additionally, predictive models for Q based on material composition are evaluated, and the potential of machine learning techniques for improving prediction accuracy is explored. The study also examines the use of ferrite microhardness and tensile strength in calculating Q and assessing thermal embrittlement. The findings provide insights for developing advanced prediction models for the thermal embrittlement behavior of CASS and the weldments of austenitic steels, contributing to the safety and reliability of nuclear power plant components.

This study investigates the effects of temperatures and strain rates on the strength and ductility of Gr.91 (ASME Grade 91) steel which is widely being used as a heat-resistant material in Generation IV nuclear and super critical thermal power plants. The tensile behavior of modified 9Cr-1Mo (Gr.91) steel was studied for the three strain rates of 6.67×10^-5/s, 6.67×10^-4/s and 6.67×10-3/s over the temperature range from room temperature (RT) to 650℃. Experimental results showed that at specific combinations of temperatures (300~400℃) and strain rates, serrations appeared in the stress-strain curves. Concurrently, abnormal behaviors such as a plateau in yield strength and tensile strength, a minimum in ductility and negative strain rate sensitivity were observed. These phenomena were analyzed as significant characteristics of dynamic strain aging (DSA). Since this abnormal behavior in Gr.91 steel affects the material strength, it is judged that a correlation analysis between DSA and material strength should be crucial in the design and integrity evaluation of Gr. 91 steel pressure vessel and piping subjected to high-temperature loading.

In this study, high temperature design and integrity evaluation methodology have been developed for Type 316L stainless steel air-to-sodium heat exchanger which uses 700℃ sodium as coolant. Currently the only design rules that take creep effects into consideration explicitly for the 316L stainless steel subjected to high temperature in the creep range are French RCC-MRx, where elevated temperature designs are possible around 550℃. Absent design coefficients at high temperature were determined based on the material properties newly determined in previous studies, and high-temperature design evaluation methodologies were developed based on 3D finite element analyses on the 700℃ class heat exchanger. In addition, works were conducted on the web-based design evaluation program of HITEP_RCC-MRx including incorporation of material properties and design coefficients up to 700℃. Methodologies on high temperature design evaluations on Type 316L stainless steel high-temperature heat exchanger were suggested.