• 대한인간공학회:학술대회논문집
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• 대한인간공학회 1993년도 추계학술대회논문집
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• pp.21-30
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• 1993

This paper describes the development of a simulation model of nuclear power plant operators including cognitive aspects by using a network modeling soft ware, Micro-SAINT (System Analysis of Integrated Networks of Tasks) for the analysis of operator performance. Network model description based on Micro-SAINT includes tasks, resources, precedence relations among tasks, flow of information and PSFs (Performance Shaping Factors) on task performance. We have tried to evaluate the performance with several performance measures such as the number of tasks allocated, relative time presure among operators within a shift, for the selected test accident scenarior; small-break LOCA (Loss of Coolant Accident) in a PWR (Pressurized Water Reactor) type nuclear power plant.

• Nuclear Engineering and Technology
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• 제28권3호
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• pp.247-256
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• 1996

This paper presents a new approach for assessing accident management strategies using containment event trees (CETs) developed during an individual plant examination (IPE) for a reference plant (CE type, 950 MWe PWR). Various accident management strategies to reduce risk have been proposed through IPE. Three strategies for the station blackout sequence are used as an example : 1) reactor cavity flooding only, 2) primary system depressurization only, and 3) doing both. These strategies are assumed to be initiated at about the time of core uncovery. The station blackout (SBO) sequence is selected in this paper since it is identified as one of the most threatening sequences to safety of the reference plant. The effectiveness and adverse effects of each accident management strategy are considered synthetically in the CETs. A best estimate assessment for the developed CETs using data obtained from NUREG-1150, other PRA results, and the MAAP code calculations is performed. The strategies are ranked with respect to minimizing the frequencies of Various containment failure modes. The proposed approach is demonstrated to be very flexible in that it can be applied to any kind of accident management strategy for any sequence.

• 설비공학논문집
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• 제29권10호
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• pp.538-550
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• 2017

In a PWR (Pressurized Water Reactor), the appropriate heat removal from the surface of fuel rod bundle is important for ensuring thermal margins and safety. Although many CFD (Computational Fluid Dynamics) software have been used to predict complex flows inside fuel assemblies with mixing vanes, there is no domestic regulatory guideline for the comprehensive evaluation of CFD software. Therefore, from the nuclear regulatory perspective, it is necessary to perform the systematic assessment and prepare the domestic regulatory guideline for checking whether valid CFD software is used for nuclear safety problems. In this study, to provide systematic evaluation and guidance on the applicability of CFD software to the domestic nuclear safety area, the results of the sensitivity analysis for the effect of the discretization scheme accuracy for the convection terms and turbulence models, which are main factors that contribute to the uncertainty in the calculation of the nuclear safety problems, on the prediction performance for the turbulent flow distribution inside the fuel assembly with split-type mixing vanes were explained.

• 소성∙가공
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• 제23권7호
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• pp.405-412
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• 2014

As a major type of heat exchanger, the steam generator (SG) produces steam from heat energy of a nuclear power plant reactor. The steam produced by the steam generator flows into a turbine, and plays an important role in electric power generation. The heat transfer tubes in the steam generator consist of approximately 10,000 U-shaped tubes, which perform a structural role and act as thermal boundaries. The heat transfer tubes conduct the thermal energy between the primary coolant (about $320^{\circ}C$, $157kgf/cm^2$) obtained from the reactor and the secondary coolant (about $260^{\circ}C$, $60kgf/cm^2$) as part of the secondary system. Recently, the heat transfer tubes in the steam generator of the pressurized water reactor (PWR) are primarily produced from Alloy 600 and Alloy 690 seamless tubes. As a pilot study to find process parameters for the cold U-bending process using rotary draw bending, numerical and experimental investigations were conducted to produce U-shaped tubes from long straight SUS304L seamless tubes. 3D finite element simulations were run using ABAQUS Explicit with consideration of the elastic recovery. The process parameters studied were the angular speed, the operation period and the bending angle. Experimental verifications were conducted to insure the suitability of the final U-shaped configurations with respect to both ovality and wall thickness.

• 한국전산구조공학회논문집
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• 제25권1호
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• pp.37-49
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• 2012

가압경수로(PWR)에서 배출되는 고준위폐기물을 지하 500m의 화강암 암반의 처분장에 장기간(약 10,000년 동안) 처분하기 위하여 여러 구조적 안전성 평가수행을 통하여 처분용기모델이 개발되었다. 기존에 설계 개발된 가압경수로용 처분용기 모델은 구조적으로 처분용기 내부에 정사각형 단면의 네 개의 고준위폐기물다발이 처분용기 단면의 중심에 대칭되게 나란히 배열된 형태를 취하고 있다. 그러나 이와 같은 배열형태가 최적의 구조인지는 아직 결정할 수 없다. 특히 경량화하는 데에는 여전히 문제가 있다. 이러한 문제를 해결하는 방법은 처분용기 단면 중심에 대하여 대칭으로 배열된 네 개의 고준위폐기물다발의 단면형상을 변경시키는 것이다. 단면형상을 변경시키는 방법에는 정사각형 형상을 유지시키면서 단면을 회전시키는 방법과 정사각형 형상을 다른 단면형상으로 변경시키는 두 가지 방법이 있다. 기 수행된 연구를 통하여 정사각형 단면형상을 유지시키면서 단면을 회전시키면 회전각도가 $30{\sim}35^{\circ}$인 배열구조의 처분용기가 나란한 정사각형 배열구조보다 구조적으로 더 안정적이어서 경량화할 수 있음을 알 수 있었다. 그러나 이 회전한 배열구조의 처분용기가 최적인지는 역시 아직 결정할 수 없다. 왜냐하면 정사각형이 아닌 다른 단면형상의 구조물에 대해서는 아직 구조적으로 더 안정한지가 확인되지 않았기 때문이다. 따라서 처분용기 단면 중심면에 대하여 대칭성을 유지하면서 고준위폐기물다발의 단면형상이 정사각형이 아닌 다른 단면형상의 처분용기구조에 대한 구조해석이 필요하다. 본 연구에서는 네 개의 고준위폐기물다발이 처분용기 중심 면에 대하여 대칭적으로 배열되면서 단면형상이 여러 가지로 변화된 가압경수로용 처분용기에 대하여 구조해석을 수행하였다. 구조해석을 수행한 결과 기존의 설계 개발된 처분용기 단면의 중심에 대칭되게 나란히 고준위폐기물다발이 배열된 정사각형 단면의 처분용기보다 다발의 단면형상이 원형인 처분용기가 구조적으로 좀 더 안정성이 있음이 밝혀졌다.

• 대한기계학회논문집A
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• 제32권2호
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• pp.177-185
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• 2008

Low-cycle environmental fatigue tests of cast austenitic stainless steel CF8M at the condition of fatigue strain rate 0.04%/sec were conducted at the pressure and temperature, 15MPa, $315^{\circ}C$ of a operating pressurized water reactor (PWR). The used test rig was limited to install an extensometer at the gauge length of the cylindrical fatigue specimen inside a small autoclave. So the magnet type LVDT#s were used to measure the fatigue displacement at the specimen shoulders inside the high temperature and high pressure water autoclave. However, the displacement and strain measured at the specimen shoulders is different from the one at the gauge length for the geometry and the cyclic strain hardening effect. Displacement of the fatigue specimen gauge length calculated by FEM (finite element method) used to modify the measured displacement and fatigue life at the shoulders. A series of low cycle fatigue life tests in air and PWR conditions simulating the cyclic strain hardening effect verified that the FEM modified fatigue life was well agreed with the simulating test results. The process and method developed in this study for the environmental fatigue test inside the small sized autoclave would be so useful to produce reliable environmental fatigue curves of CF8M stainless steel in pressurized water reactors.

• Corrosion Science and Technology
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• 제20권3호
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• pp.158-168
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• 2021

Austenitic 316 stainless steel was irradiated with protons accelerated by an energy of 2 MeV at 360 ℃, the various defects induced by this proton irradiation were characterized with microscopic equipment. In our observations irradiation defects such as dislocations and micro-voids were clearly revealed. The typical irradiation defects observed differed according to depth, indicating the evolution of irradiation defects follows the characteristics of radiation damage profiles that depend on depth. Surface oxidation tests were conducted under the simulated primary water conditions of a pressurized water reactor (PWR) to understand the role irradiation defects play in surface oxidation behavior and also to investigate the resultant irradiation assisted stress corrosion cracking (IASCC) susceptibility that occurs after exposure to PWR primary water. We found that Cr and Fe became depleted while Ni was enriched at the grain boundary beneath the surface oxidation layer both in the non-irradiated and proton-irradiated specimens. However, the degree of Cr/Fe depletion and Ni enrichment was much higher in the proton-irradiated sample than in the non-irradiated one owing to radiation-induced segregation and the irradiation defects. The microstructural and microchemical changes induced by proton irradiation all appear to significantly increase the susceptibility of austenitic 316 stainless steel to IASCC.

• Nuclear Engineering and Technology
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• 제23권1호
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• pp.66-80
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• 1991

전형적 PWR 비상노심냉각계통에서의 12 inch 파단사고에 대응하는 LOFT 중형냉각재 상실사고 모사 실험 자료 L5-1을 이용하여 RELAP5/MOD2 Cycle 36.04 전산코드의 평가가 수행되었다. 평가 근거는 기준 코드와 nodalization에 의한 계산 결과가 L5-1 실험 결과와 잘 일치하는지, 추가적인 민감도 분석 연구로는 이중 노심 유로 및 열적 모델을 고려하고 model 민감도 분석으로는 reflood, gap conductance option 사용 여부에 따른 피복재 온도에 미치는 영향을 관찰하였다. 기준 계산 결과 기준 모델이 L5-1 현상을 대체로 잘 모사하였으나, 피복재가 천천히 가열되고 주변 부위의 피복재 온도가 과대하게 예견되었다. 민감도 분석 결과 단일 열적 모델이 피복재 가열 시작을 10초개선 하였고, 이중 유로 모델이 주변 온도를 20K 개선하였으나 최대 피복재 온도는 기준 계산시 보다 정확치 못하였으므로, 기준 모델인 단일 유로, 이중 열적 구조 그리고 reflood option은 사용하고 gap conductance option은 사용치 않는 것이 코드의 중형 냉각재 상실사고 해석시 피복재 온도 관찰의 관점에서 바람직하다.

• Nuclear Engineering and Technology
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• 제41권2호
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• pp.187-198
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• 2009

In this paper, three power ramp tests performed on high burn-up Re-crystallized Zircaloy2 - UO2 BWR fuel rods (56 to 63 MWd/kgU) within the SCIP project are simulated with METEOR and ALCYONE 3D. Two of the ramp tests are of staircase type up to Linear Heat Rates of 420 and 520 W/cm and with long holding periods. Failure of the 420 W/cm fuel rod was observed after 40 minutes. The third ramp test consisted of a more standard ramp test with a constant power rate of 80 W/cm/min up to 410 W/cm with a short holding time. The tests were first simulated with the METEOR 1D fuel rod code, which gave accurate results in terms of profilometry and fission gas releases. The behaviour of a fuel pellet fragment and of the cladding piece on top of it was then investigated with ALCYONE 3D. The size and the main characteristics of the ridges after base irradiation and power ramp testing were recovered. Finally, the failure criteria validated for PWR conditions and fuel rods with low-to-medium burn-ups were used to analyze the failure probability of the KKL rodlets during ramp testing.

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2002년도 유체기계 연구개발 발표회 논문집
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• pp.493-498
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• 2002

The letdown system of pressurized water reactor (PWR) nuclear fewer plants had experienced instabilities in letdown system due to unacceptable flow characteristics of control valves. The Korean Standard Nuclear Power Plants (KSNPs) have three flow paths in parallel for letdown new control. Each flow path consists of two offices and one isolation valve. This study evaluates the effect of orifice arrangement and valve stroke time of letdown isolation valve on the system transients because sudden flow changes due to valve actuation can generate high pressure peaks in letdown line. A pressure transient analysis has been preformed to evaluate the impact of dynamic transients. This analysis uses MMS which is a simulation code developed by EPRI based on the method of characteristics. The result shows that the pressure peak is reduced in the continuous arrangement but negligible. Additionally, it shows that the stroke time of linear type flog valve greater than 15 seconds can give more stable performance.