• Nuclear Engineering and Technology
• /
• v.54 no.3
• /
• pp.849-859
• /
• 2022

Feed-water valves (FWVs) are used to regulate the flow rate of water entering steam generators, which are very important devices in nuclear power plants. Due to the working environment of relatively high pressure and temperature, there is strength failure problem of valve body in some cases. Based on the thermo-fluid-solid coupling model, the valve body stress of the feed-water valve in the opening process is investigated. The flow field characteristics inside the valve and temperature change of the valve body with time are studied. The stress analysis of the valve body is carried out considering mechanical stress and thermal stress comprehensively. The results show that the area with relatively high-velocity area moves gradually from the bottom of the cross section to the top of the cross section with the increase of the opening degree. The whole valve body reaches the same temperature of 250 ℃ at the time of 1894 s. The maximum stress of the valve body meets the design requirements by stress assessment. This work can be referred for the design of FWVs and other similar valves.

• Nuclear Engineering and Technology
• /
• v.55 no.9
• /
• pp.3342-3351
• /
• 2023

High-density polyethylene (HDPE) pipelines in nuclear power plants (NPPs) have to meet high requirements for seismic performance. HDPE pipes have been proved to have good seismic performance, but joints are the weak links in the pipelines, and pipeline failures usually initiate from the defects inside the joints. Limited data are available on the seismic performance of butt-fusion joints of HDPE pipelines in NPPs, especially in terms of defects changes inside the joints after earthquakes. In this paper, full-scale shaking table tests were performed on a test section of suspended HDPE pipelines in an NPP, which included straight pipes, elbows, and 10 butt-fusion joints. During the tests, the seismic load-induced strain of the joints was analyzed by strain gauges, and it was much smaller than the internal pressure and self-weight-induced strain. Before and after the shaking table tests, phased array ultrasonic testing (PA-UT) was conducted to detect defects inside the joints. The locations, numbers, and dimensions of the defects were analyzed. It was found that defects were more likely to occur in elbows joints. No new defect was observed after the shaking table tests, and the defects showed no significant growth, indicating the satisfactory seismic performance of the butt-fusion joints.

• Proceedings of the KIPE Conference
• /
• 2018.07a
• /
• pp.177-179
• /
• 2018

After Fukushima nuclear accident at 2011, nuclear industrial has been focused on operation and maintenance phase, not design and construction phase. Continued good operating performance of nuclear power plants has been the best critical issue to nuclear utilities. Replacement for complete components as well as parts of components is being procured because nuclear utilities must maintain safety and reliability of operating nuclear power plants. However, many suppliers and manufacturers are giving up a nuclear quality assurance program under reduction in new construction of nuclear power plants. It is able to be increased difficulty in procuring spare parts to support operations and maintenance of nuclear power plants. Over 20% of nuclear power plant equipment in some countries is obsolete. Owing to obsolescence of nuclear safety-related items and/or withdrawing a nuclear quality assurance program of suppliers and manufactures, some replacement item and part might be procured to the item not covered by appendix B to USNRC 10 CFR Part 50. Under various methods of the nuclear repair and replacement methodology, utilities are supposed to establish a typical program for a repair and replacement of an electrical equipment and its parts in conjunction with a nuclear quality assurance. Concerning this typical program, this study suggests the repair and replacement methodology of electrical equipments used in nuclear power plants by procurement of a power supply, based on nuclear regulations, codes, standards, guidelines, specific and general technical information, etc..

• Nuclear Engineering and Technology
• /
• v.52 no.8
• /
• pp.1661-1668
• /
• 2020

A physical protection system (PPS) is used for the protection of critical facilities. This paper proposes a structure analytic hierarchy approach (SAHA) for the hierarchical evaluation of the PPS effectiveness in critical infrastructure. SAHA is based on the traditional analysis methods "estimate of adversary sequence interruption, EASI". A community algorithm is used in the building of the SAHA model. SAHA is applied to cluster the associated protection elements for the topological design of complicated PPS with graphical vertexes equivalent to protection elements.

• Nuclear Engineering and Technology
• /
• v.53 no.11
• /
• pp.3474-3490
• /
• 2021

Sodium-cooled fast reactor (SFR) is the preferred technology of the generation-IV fast neutron reactor, and its core body mainly uses nuclear-grade 316 stainless steel. In order to prolong the design life of SFRs to 60 years and more, it is necessary to summarize and analyze the anti-corrosion effect of nuclear grade 316 stainless steel in high temperature sodium environment. The research on sodium corrosion of nuclear grade 316 stainless steel is mainly composed of several important factors, including the microstructure of stainless steel (ferrite layer, degradation layer, etc.), the trace chemical elements of stainless steel (Cr, Ni and Mo, etc) and liquid impurity elements in sodium (O, C and N, etc), carburization and mechanical properties of stainless steel, etc. Through summarizing and constructing the sodium corrosion rate equations of nuclear grade 316 stainless steel, the stainless steel loss of thickness can be predicted. By analyzing the effects of temperature, oxygen content in sodium and velocity of sodium on corrosion rate, the basis for establishing integrity evaluation standard of SFR core components with sodium corrosion is provided.

• The Korean Journal of Nuclear Medicine Technology
• /
• v.28 no.1
• /
• pp.71-79
• /
• 2024

Purpose: This study aims to identify SUV, SNR, spatial resolution, and axial uniformity under the same reconstruction conditions and to find out the differences between equipment models. Materials and Methods: The equipment was GE's Discovery 600, 710, IQ, MI(GE Healthcare, USA), and the Phantom used ACR(American College of Radiology) Flangeless Esser Phantom and PET/SPECT Performance Phantom. The PET/SPECT Performance Phantom injected 18F-FDG at a concentration of 3.8 kBq/mL, and the ACR Flangeless Esser Phantom made the conditions for Hot Spot and Background activity for 4 : 1. Image evaluation was compared and evaluated for SUV, SNR, spatial resolution, and axial uniformity with the same reconstruction that added SharpIR of VPHD. Results: The SUV_max showed a difference up to 4.6% with an average of 2.71, 2.35, 1.89, and 1.43 from Hot Spot 1 to 4, and the SUV_mean showed a difference up to 4.7% with 2.06, 1.75, 1.49, and 1.27. There was a difference up to 5% between equipment, and there was no significant difference between both SUV_max and SUV_mean. SNR showed a difference up to 0.04 with an average of 0.37, 0.26, 0.18, and 0.11. FWHM showed a difference up to 0.27. Lastly, COV of axial uniformity was up to 0.018. Conclusion: SUV showed differences within 5% between equipment and showed no significant difference. This is considered to be used as basic data that can be used for the development and replacement of equipment because it has the advantage of being able to observe with a large number of equipment.

• Proceedings of the KIEE Conference
• /
• 2007.04a
• /
• pp.143-145
• /
• 2007

The development of excitation control system (ECS) for emergency diesel generator in nuclear power plant is the replacement project of existing control system to resolve the maintenance problems caused by aging and obsolescence, The excitation control system is classified as a safety-related system. To guarantee the performance of developing excitation control system is equal to or higher than that of other systems, establishing the quality assurance scheme, doing software verification and validation activities, and planning equipment qualification. In this paper, we'd like to introduce the equipment qualification of excitation control system.

• Journal of Korean Society of Industrial and Systems Engineering
• /
• v.25 no.3
• /
• pp.1-8
• /
• 2002

This paper investigates different QA standards such as KEPIC QAP, KEPIC END 1200, ISO/1EC 17025 etc. and as a result defines QA elements for Nuclear Power Plant equipment qualification(EQ) in Korea. This paper also proposes a practical QA certification system appropriate for an Integrated Organization for EQ which is being planned to be established in Korea. Since the level of the Korean EQ technology is comparatively low, the Korean manufacturers of the Nuclear Power Plant(NPP) equipment have usually used overseas EQ services. The EQ related organizations in Korea are making efforts to construct the integrated EQ system. In connection with this, it is required that the QA elements and QA certification system suitable for EQ in Korea be developed.

• Nuclear Engineering and Technology
• /
• v.49 no.5
• /
• pp.914-919
• /
• 2017

While industry understands the importance of keeping equipment operational and well maintained, the importance of tracking maintenance information in reliability models is often overlooked. Prognostic models can be used to predict the failure times of critical equipment, but more often than not, these models assume that all maintenance actions are the same or do not consider maintenance at all. This study investigates the influence of integrating maintenance information on prognostic model prediction accuracy. By incorporating maintenance information to develop maintenance-dependent prognostic models, prediction accuracy was improved by more than 40% compared with traditional maintenance-independent models. This study acts as a proof of concept, showing the importance of utilizing maintenance information in modern prognostics for industrial equipment.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.05a
• /
• pp.1112-1113
• /
• 2005