• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1874-1879
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• 2024

Owing to strict radiation safety management in Korean nuclear power plants (NPPs), most radiation workers receive very low radiation doses, even lower than the annual dose limit for the general public. However, the occupational dose distribution indicates that some Korean NPP workers receive a relatively higher dose than the average dose. This inequity in radiation exposure could be reduced by providing customized radiation protection measures, such as dose constraints, to workers receiving relatively higher doses. In this study, dose normalization was performed to identify the highest radiation exposure work in Korean pressurized water reactors (PWRs). The results show that most of the occupational exposure in Korean PWRs occurs during the planned maintenance period. Finally, the three highest radiation exposure tasks in Korean PWRs were identified: nozzle dam installation and removal, eddy current testing, and man-way opening and closing.

• Nuclear Engineering and Technology
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• 제42권5호
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• pp.582-589
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• 2010

As a part of the back-end fuel cycle, transportation of spent nuclear fuel (SNF) from nuclear power plants (NPPs) to a fuel storage facility is very important in establishing a nuclear fuel cycle. In Korea, the accumulated amount of SNF in the NPP pools is troublesome since the temporary storage facilities at these NPP pools are expected to be full of SNF within ten years. Therefore, Korea cannot help but plan for the construction of an interim storage facility to solve this problem in the near future. Especially, a decision on several factors, such as where the interim storage facility should be located, how many casks a transport ship can carry at a time and how many casks are initially required, affect the configuration of the transportation system. In order to analyze the various possible candidate scenarios, we assumed four cases for the interim storage facility location, three cases for the load capacity that a transport ship can carry and two cases for the total amount of casks used for transportation. First, this study considered the currently accumulated amount of SNF in Korea, and the amount of SNF generated from NPPs until all NPPs are shut down. Then, how much SNF per year must be transported from the NPPs to an interim storage facility was calculated during an assumed transportation period. Second, 24 candidate transportation scenarios were constructed by a combination of the decision factors. To construct viable yearly transportation schedules for the selected 24 scenarios, we created a spreadsheet program named TranScenario, which was developed by using MS EXCEL. TranScenario can help schedulers input shipping routes and allocate transportation casks. Also, TranScenario provides information on the cask distribution in the NPPs and in the interim storage facility automatically, by displaying it in real time according to the shipping routes, cask types and cask numbers that the user generates. Once a yearly transportation schedule is established, TranScenario provides some statistical information, such as the voyage time, the availability of the interim storage facility, the number of transported casks sent from the NPPs, and the number of transported casks received at the interim storage facility. By using this information, users can verify and validate a yearly transportation schedule. In this way, the 24 candidate scenarios could be constructed easily. Finally, these 24 scenarios were compared in terms of their operation cost.

• 한국압력기기공학회 논문집
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• 제11권2호
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• pp.31-39
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• 2015

Pipe wall-thinning by flow-accelerated corrosion (FAC) and various types of erosion is a significant and costly damage phenomenon in secondary piping systems of nuclear power plants (NPPs). Most NPPs have management programs to ensure pipe integrity due to wall-thinning that includes periodic measurements for pipe wall thicknesses using ultrasonic tests (UTs). Nevertheless, thinning evaluations are not easy because the amount of thickness reduction being measured is often quite small compared to the accuracy of the inspection technique. U.S. Electric Power Research Institute (EPRI) had proposed Total Point Method (TPM) as a thinning occurrence evaluation method, which is a very useful method for detecting locally thinned pipes or fittings. However, evaluation engineers have to discern manually the measurement data because there are no numerical algorithm for TPM. In this study, numerical algorithms were developed based on non-parametric and parametric statistical method.

• 한국압력기기공학회 논문집
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• 제15권1호
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• pp.53-61
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• 2019

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.

• 대한인간공학회지
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• 제22권1호
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• pp.31-42
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• 2003

This study describes a modification of the technique for human error analysis in nuclear power plants (NPPs) which adopts advanced Man-Machine Interface (MMI) features based on computerized working environment, such as LCOs. Flat Panels. Large Wall Board, and computerized procedures. Firstly, the state of the art on human error analysis methods and efforts were briefly reviewed. Human error analysis method applied to NPP design has been THERP and ASEP mainly utilizing Swain's HRA handbook, which has not been facilitated enough to put the varied characteristics of MMI into HRA process. The basic concepts on human errors and the system safety approach were revisited, and adopted the process of FMEA with the new definition of Error Segment (ESJ. A modified human error analysis process was suggested. Then, the suggested method was applied to the failure of manual pump actuation through LCD touch screen in loss of feed water event in order to verify the applicability of the proposed method in practices. The example showed that the method become more facilitated to consider the concerns of the introduction of advanced MMI devices, and to integrate human error analysis process not only into HRA/PRA but also into the MMI and interface design. Finally, the possible extensions and further efforts required to obtain the applicability of the suggested method were discussed.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 2006년도 추계 학술발표회 논문집
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• pp.237-240
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• 2006

Almost 30 years have been passed since the first nuclear power plant was operated in Korea. Many studies have been actively conducted from the early 1990's in order to develop the deterioration management system for concrete structures in NPPs(Nuclear Power Plants) accordingly. Base on these studies, a systematic deterioration management program has developed and operated since 1997. According to this program, systematic inspections to provide database and evaluation were periodically performed (every overhaul at intervals of $12{\sim}18$ month and every five years). Accumulated deterioration database was usefully utilized for the NPP PSR (Periodic Safety Review). In this paper, the long-term durability and integrity of Kori 1,2 NPP concrete structures which are the oldest ones in Korea were evaluated based on the precise inspection database and regulatory inspection results including compressive strength, depth of carbonation, amount of chlorination and spontaneous potential of reinforcing bar, etc. It was noted that Kori 1,2 NPP structures have not any serious durability problems.

• Nuclear Engineering and Technology
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• 제52권2호
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• pp.248-257
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• 2020

This study presents an initial design for a novel system consisting in a coupled nuclear reactor and a phase change material-based thermal energy storage (TES) component, which acts as a buffer and regulator of heat transfer between the primary and secondary loops. The goal of this concept is to enhance the capacity factor of nuclear power plants (NPPs) in the case of high integration of renewable energy sources into the electric grid. Hence, this system could support in elevating the economics of NPPs in current competitive markets, especially with subsidized solar and wind energy sources, and relatively low oil and gas prices. Furthermore, utilizing a prismatic-core advanced high temperature reactor (PAHTR) cooled by a molten salt with a high melting point, have the potential in increasing the system efficiency due to its high operating temperature, and providing the baseline requirements for coupling other process heat applications. The present research studies the neutronics and thermal hydraulics (TH) of the PAHTR as well as TH calculations for the TES which consists of 300 blocks with a total heat storage capacity of 150 MWd. SERPENT Monte Carlo and MCNP5 codes carried out the neutronics analysis of the PAHTR which is sized to have a 5-year refueling cycle and rated power of 300 MW_th. The PAHTR has 10 metric tons of heavy metal with 19.75 wt% enriched UO₂ TRISO fuel, a hot clean excess reactivity and shutdown margin of $33.70 and -$115.68; respectively, negative temperature feedback coefficients, and an axial flux peaking factor of 1.68. Star-CCM + code predicted the correct convective heat transfer coefficient variations for both the reactor and the storage. TH analysis results show that the flow in the primary loop (in the reactor and TES) remains in the developing mixed convection regime while it reaches a fully developed flow in the secondary loop.

• 비파괴검사학회지
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• 제33권2호
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• pp.147-153
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• 2013

최근 가압경수로형 원전의 가압기 노즐과 안전단 사이의 이종용접부에서 일차수응력부식균열에 대한 건전성 확보가 중요한 관심사항으로 대두되고 있다. 가압기 노즐은 SA508 Gr.3 저합금강이며 안전단은 F316L 스테인리스강으로, 이들 두 재료 사이에 용접재로는 Alloy 82/182가 사용되었다. 재료 결함에 대한 건전성 평가를 위해서는 재료의 기계적 물성치, 특히 인장물성과 파괴물성이 확보되어야 한다. 그러나, 일반적인 재료 규격과 시험성적서에서는 상온의 인장물성이 제공되지만 고온의 인장물성과 파괴인성이 제공되지 않는다. 따라서, 본 논문에서는 상온과 원전 운전온도에서 SA508 Gr.3과 F316L 스테인리스강에 대한 인장시험과 J-R 파괴인성시험을 수행하고 그 결과를 수록하였다.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.386-400
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• 2022

Seismic fragility analysis, a part of seismic probabilistic risk assessment (SPRA), is commonly used to establish the relationship between a representative property of earthquakes and the failure probability of a structure, component, or system. Current guidelines on the SPRA of nuclear power plants (NPPs) used worldwide mainly reflect the earthquake characteristics of the western United States. However, different earthquake characteristics may have a significant impact on the seismic fragility of a structure. Given the concern, this study aimed to investigate the effects of earthquake characteristics on the seismic fragility of concrete containments housing the OPR-1000 reactor. Earthquake time histories were created from 30 ground motions (including those of the 2016 Gyeongju earthquake) by spectral matching to the site-specific response spectrum of Hanbit nuclear power plants in South Korea. Fragility curves of the containment structure were determined under the linear response history analysis using a lumped-mass stick model and 30 ground motions, and were compared in terms of earthquake characteristics. The results showed that the median capacity and high confidence of low probability of failure (HCLPF) tended to highly depend on the sustained maximum acceleration (SMA), and increase when using the time histories which have lower SMA compared with the others.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.223-229
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• 1995

Loose Par Monitoring System(LPMS) is one of the fundamental diagnostic tools installed in the nuclear power plants. In this paper, recovery process algorithm and model for the corrupted impact signal generated by loose parts is presented. The characteristics of this algorithm can obtain a proper burst signal even though background noise is considerably high level comparing with actual impact signal. To verify performance of the proposed algorithm, we evaluate mathematically signal-to-noise ratio of primary output and noise. The performance of this recovery process algorithm is shown through computer simulation.