• 한국압력기기공학회 논문집
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• 제15권2호
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• pp.40-49
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• 2019

In this paper, failure probabilities of the OPR1000 reactor vessel under pressurized thermal shock (PTS) were estimated using the probabilistic fracture mechanics code, R-PIE. Input variables of initial crack distribution, crack size, copper contents, and upper shelf toughness were selected for the sensitivity analyses. A wide range of the input data were considered. Through-wall cracking frequencies determined by the product of the vessel failure probability and the corresponding occurrence frequency of the transient were also compared to the acceptance criterion. The results showed that transient history had the most significant impact on the vessel failure probability. Moreover, conservative assumptions resulted in extremely high through-wall cracking frequencies.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.1167-1180
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• 2023

Due to the tightened regulatory environment since the Fukushima accident, the capacity factor of Korean nuclear power plants has been declining since 2011. To overcome this circumstance, a shift from 18-month to 24-month cycle operation is being considered in Korea. Therefore, in this study, loading patterns(LPs) for 24-month cycle operation of the Korean standard nuclear power plant(OPR-1000) are suggested and economic evaluations are performed. A single-zone LP with 89 fresh fuels was evaluated to be optimal for 24-month operation of OPR-1000 in terms of economic gain. The 24-month operation of OPR-1000 with this LP gives a profit of 7.073 million dollars per year compared to 18-month operation.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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• pp.287-292
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• 2008

The applicability of stainless steel type 316N to the PZR surge-lines of OPR1000 and APR1400 is investigated. So far, strainless steel type 347 has been used for the OPR1000 surge-lines. The degree of improvement in the leak-before-break(LBB) and component design margin is evaluated when stainless steel type 347 is substituted by type 316N. For the study, the tensile and J-R tests on type 316N and type 347 stainless steels were performed at 316 and the microstructure of both types was examined. Stainless steel type 316N shows the higher values on the stress-strain curves, J-R curves and stress intensity, Sm, compared to those of type 347. Therefore, stainless steel type 316N ensures the higher LBB and component design margins. As a result, this study shows that stainless steel type 316N could substitute type 347 for the surge-lines of OPR1000 and APR1400.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2013년도 추계학술대회 논문집
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• pp.402-403
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• 2013

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.318-329
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• 2016

Reducing critical boron concentration in a commercial pressurized water reactor core offers many advantages in view of safety and economics. This paper presents a preliminary investigation of a reduced-boron pressurized water reactor core to achieve a clearly negative moderator temperature coefficient at hot zero power using the newly-proposed "Burnable absorber-Integrated Guide Thimble" (BigT) absorbers. The reference core is based on a commercial OPR1000 equilibrium configuration. The reduced-boron ORP1000 configuration was determined by simply replacing commercial gadolinia-based burnable absorbers with the optimized BigT-loaded design. The equilibrium cores in this study were directly searched via repetitive Monte Carlo depletion calculations until convergence. The results demonstrate that, with the same fuel management scheme as in the reference core, application of the BigT absorbers can effectively reduce the critical boron concentration at the beginning of cycle by about 65 ppm. More crucially, the analyses indicate promising potential of the reduced-boron OPR1000 core with the BigT absorbers, as its moderator temperature coefficient at the beginning of cycle is clearly more negative and all other vital neutronic parameters are within practical safety limits. All simulations were completed using the Monte Carlo Serpent code with the ENDF/B-VII.0 library.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2812-2820
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• 2020

House load operation (HLO) occurs when the generator supplies power to the house load without triggering reactor trips during grid disturbances. In Korea, the HLO capability of optimized power reactor 1000 (OPR1000) plants has prevented several reactor trips. Operational experiences demonstrate the difference in the reactor trip incidence due to grid disturbances between OPR1000 plants and Westinghouse plants in Korea, attributable to the availability of the HLO capability. However, probabilistic safety assessments (PSAs) for OPR1000 plants have not considered their specific design features in the initiating event analyses. In an at-power PSA, the HLO capability can affect the initiating event frequencies of general transients (GTRN) and loss of offsite power (LOOP), resulting from transients within the grid system. The initiating event frequencies of GTRN and LOOP for an OPR1000 plant are reduced by 17.7% and 78.7%, respectively, compared to the Korean industry-average initiating event frequencies, and its core damage frequency from internal events is reduced by 15.2%. The explicit consideration of the HLO capability in initiating event analyses makes significant changes in the risk contributions of the initiating events. Consequently, for more realistic at-power PSAs in Korea, we recommend incorporating plant-specific HLO-related design features when estimating initiating event frequencies.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1513-1525
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• 2024

An improved mitigation system for thermally-induced steam generator tube rupture accidents was introduced to prevent direct environmental release of fission products bypassing the containment in the OPR1000. This involves injecting bypassed steam into the containment, cooling, and decontaminating it using a water coolant tank. To evaluate its performance, a severe accident analysis was performed using the MELCOR 2.2 code for OPR1000. Simulation results show that the proposed system sufficiently prevented the release of radioactive nuclides (RNs) into the environment via containment injection. The pool scrubbing system effectively decontaminated the injected RN and consequently reduced the aerosol mass in the containment atmosphere. However, the decay heat of the collected RNs causes re-vaporization. To restrict the re-vaporization, an external water source was considered, where the decontamination performance was significantly improved, and the RNs were effectively isolated. However, due to the continuous evaporation of the feed water caused by decay heat, a substantial amount of steam is released into the containment. Despite the slight pressurization inside the containment by the injected and evaporated steam, the steam decreased the hydrogen mole fraction, thereby reducing the possibility of ignition.

• 한국압력기기공학회 논문집
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• 제6권2호
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• pp.16-19
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• 2010

In this study, the axial ODSCC occurrence of domestic OPR 1000 steam generator tube was caused by the tube weakness and the sludge accumulation in the secondary side of steam generator. Inconel 600 HTMA used as tube material is related to most of tube leakage accidents in the world and also these ODSCCs were detected mainly at the 5th TSP(Tube Support Plate) to the 8th TSP of hot leg side. These elevations(5th TSP to 8th TSP) pave the way for the sludge accumulation. As a result of EC(Eddy Current) Bobbin and RPC data analysis, ODSCCs were occurred at contact points of tube and tube support plate. The more accumulated sludge, the higher occurrence frequency of ODSCC.

• 한국전산구조공학회논문집
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• 제26권3호
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• pp.207-212
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• 2013

APR1000(Advanced Power Reactor 1000)은 기존의 OPR1000(Optimized Power Reactor 1000)에 60년 설계수명, 국부주파수제어운전, 0.3g 안전정지지진하중 적용 등의 향상된 설계특성(Advanced Design Feature)을 적용하여 개선한 수출형 1000MW 원전이다. 이 논문에서는 Reg. Guide 1.207에서 요구하는 원자로냉각재 환경을 고려한 피로 평가를 원자로용기에 대하여 평가하였다. 원자로용기에서 비교적 누적사용계수가 높은 출구노즐을 대상으로 평가를 수행하였으며 출구노즐은 구조적 건전성을 만족하는 것으로 평가되었다.

• 시스템엔지니어링학술지
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• 제11권1호
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• pp.1-7
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• 2015

The purpose of this paper is to present the results of reengineering of the YGN 3&4 (Yonggwang Nuclear Power Plant, Units 3&4) SDS (Safety Depressurization System) retrofit design and to make recommendations for the improvement in design and design procedure implementing the Systems Engineering (SE) process. YGN 3&4 is a basic model for OPR1000 (the Korean standard 1000 MWe plant). The basic model, herein, represents the reference plant for the OPR1000 development. In the middle of the YGN 3&4 construction, the Korean Nuclear Regulatory Body requested a retrofit of this plant with a means to rapidly depressurize the plant in conformance with a severe accident mitigation requirement. For the reengineering of the SDS in YGN 3&4, V-model and functional and physical architectures have been developed. A SE decision making method has been used for the selection of SDS valves. Finally, recommendations have been made to improve OPR1000 design for the improved operation and enhanced safety.