• 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 Over 20% of nuclear power plant equipment in some countries is obsolete.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

• Journal of the Earthquake Engineering Society of Korea
• /
• v.24 no.1
• /
• pp.1-8
• /
• 2020

In 2016, an earthquake occurred at Gyeongju, Korea. At the Wolsong site, the observed peak ground acceleration was lower than the operating basis earthquake (OBE) level of Wolsong nuclear power plant. However, the measured spectral acceleration value exceeded the spectral acceleration of the operating-basis earthquake (OBE) level in some sections of the response spectrum, resulting in a manual shutdown of the nuclear power plant. Analysis of the response spectra shape of the Gyeongju earthquake motion showed that the high-frequency components are stronger than the response spectra shape used in nuclear power plant design. Therefore, the seismic performance evaluation of structures and equipment of nuclear power plants should be made to reflect the characteristics of site-specific earthquakes. In general, the floor response spectrum shape at the installation site or the generalized response spectrum shape is used for the seismic performance evaluation of structures and equipment. In this study, a generalized response spectrum shape is proposed for seismic performance evaluation of structures and equipment for nuclear power plants. The proposed response spectrum shape reflects the characteristics of earthquake motion in Korea through earthquake hazard analysis, and it can be applied to structures and equipment at various locations.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2004.11a
• /
• pp.755-762
• /
• 2004

In our country, when the replacement for individual components of equipment in nuclear power plants is required, establishment of individual criteria i.e. Required Response Spectra(RRS) of seismic test/analysis for the component is very difficult because of the absence of Test Response Spectra(TRS) for the individual component to be replaced, from the existing qualification documents. In this case, it is required to perform the structural analysis for the nuclear equipment including the components to be replaced. After the structural analysis, Analysis Response Spectra(ARS) at the point of the component shall be generated and used for seismic test of the component. However, as of today, no standard program authorized for the response spectra generation by using the structural analysis exists in korea. Because of above reason, the STAR-Egs computer program was developed by using the method which calculates directly the expected response spectrum(frequency vs. acceleration type) of the selected points in the nuclear equipment with input spectrum(Required Response Spectra, RRS), based on the dynamic characteristics of the Finite Element(FE) model that is equivalent to the nuclear equipment. The STAR-Egs controls ANSYS/I-DEAS commercial software and automatically extract modal parameters of the FE model. The STAR-Egs calculates response spectrum using the established algorithm based on the extracted modal parameters, too. Reliance on the calculation result of the STAR-Egs was verified through comparison output with the result of MATLAB commercial software based on the identical algorithm. Moreover, actual seismic testing was performed as per IEEE344-1987 for the purpose of program verification by comparison of the FE analysis results.