• 대한기계학회논문집A
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• 제37권9호
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• pp.1159-1168
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• 2013

가압형 경수로 원자로의 압력용기 상부헤드 관통노즐 J-groove 용접부 주변에서 일차수응력부식균열(PWSCC)로 인한 냉각수 누설사례가 발생하고 있다. 본 연구에서는 PWSCC 의 주요 원인 중 하나인 용접 잔류응력을 유한요소 해석을 이용해 평가하고 원자력 발전소의 정상가동 조건을 해석에 반영하는 방법이 용접잔류응력 분포에 미치는 영향에 대한 분석을 수행하였다. 또한 반복되는 원자력 발전소의 가동 주기가 용접잔류응력 분포에 미치는 영향을 확인하여 정상가동조건에서의 정확한 용접 잔류응력을 예측할 수 있는 방법을 분석하였다.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.1085-1097
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• 2022

A printed circuit steam generator (PCSG) is being considered as the component for pressurized water reactor (PWR) type small modular reactor (SMR) that can further reduce the physical size of the system. Since a steam generator in many PWR-type SMR generates superheated steam, it is expected that dryout front oscillation can potentially cause thermal fatigue failure due to cyclic thermal stresses induced by the transition in boiling regimes between convective evaporation and film boiling. To investigate the fatigue issue of a PCSG, a reference PCSG is designed in this study first using an in-house PCSG design tool. For the stress analysis, a finite element method analysis model is developed to obtain the temperature and stress fields of the designed PCSG. Fatigue estimation is performed based on ASME Boiler and pressure vessel code to identify the major parameters influencing the fatigue life time originating from the dryout front oscillation. As a result of this study, the limit on the temperature difference between the hot side and cold side fluids is obtained. Moreover, it is found that the heat transfer coefficient of convective evaporation and film boiling regimes play an essential role in the fatigue life cycle as well as the temperature difference.

• Nuclear Engineering and Technology
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• 제25권3호
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• pp.413-423
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• 1993

본 논문에서는 경수로 노심 용융사고시 1차계통의 압력이 높은 경우에 발생하는 격납용기 직접가열 현상에 대한 실험연구를 하였다. 실험은 고리 1,2호기와 영광 3,4호기의 1/30 축소규모와 고리 1,2호기의 1/20 축소규모를 실험모형으로 하여 수행되었으며, 고리 1,2호기의 경우 축소 규모에 따른 검증도 시도하였다. 실험의 주요 변수는 초기 압력 용기의 압력, 파열면적 및 캐비티의 구조 등이다. 실험결과로부터 캐비티 외부로의 용융노심 분사비율은 높은 초기압력과 큰 파열면적을 가진 경우가 더 높으며 캐 비티의 구조가 분사비율에 큰 영향을 미침을 알 수 있었다. 본 연구의 실험결과를 이용하여 분사비율에 대한 실험관계식을 무차원 유효시간의 함수로 도출하여 제시하였으며, 이 실험관계식은 본 실험결과 뿐만 아니라 한국 과학기술원의 실험자료 및 미국 BNL 실험결과와도 잘 일치하였다.

• 한국기계가공학회지
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• 제11권4호
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• pp.160-167
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• 2012

This paper is concerned with a particular fracture mechanics parameter ${\Delta}K_{th}$, known as the 'threshold stress intensity range', or 'fatigue threshold'. This threshold ${\Delta}K_{th}$ constitutes, as it were, a hinge between the notion of crack initiation and the notion of crack growth. It has often been thought that, like the endurance limit, it could be an intrinsic criterion of the material. The study was conducted on a SA-508 pressure vessel steel used in the nuclear power industry. This material exhibits a typical threshold effect in the range of the crack growth rates which were determined; that is, below approximately $da/dN=10^{-6}mm/cycle$, the slope of the da./dN versus ${\Delta}K$ curve is almost vertical. The value of ${\Delta}K_{th}$ was determined at a growth rate of $10^{-7}$ mm/cycle according to the ASTM Standard for threshold testing. The fatigue threshold values are in the range 21 $kg/mm^{3/2}$ to 12 $kg/mm^{3/2}$ depending on the stress ratio effect.

• 대한금속재료학회지
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• 제48권3호
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• pp.194-202
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• 2010

Application of a stronger and more durable material for reactor pressure vessels (RPVs) might be an effective way to insure the integrity and increase the efficiency of nuclear power plants. A series of research projects to apply the SA508 Gr.4 steel in ASME code to RPVs are in progress because of its excellent strength and durability compared to commercial RPV steel (SA508 Gr.3 steel). In this study, the microstructural characteristics and mechanical properties of SA508 Gr.3 Mn-Mo-Ni low alloy steel and SA508 Gr.4N Ni-Mo-Cr low alloy steel were investigated. The differences in the stable phases between these two low alloy steels were evaluated by means of a thermodynamic calculation using ThermoCalc. They were then compared to microstructural features and correlated with mechanical properties. Mn-Mo-Ni low alloy steel shows the upper bainite structure that has coarse cementite in the lath boundaries. However, Ni-Mo-Cr low alloy steel shows the mixture of lower bainite and tempered martensite structure that homogeneously precipitates the small carbides such as $M_{23}C_6$ and $M_7C_3$ due to an increase of hardenability and Cr addition. In the mechanical properties, Ni-Mo-Cr low alloy steel has higher strength and toughness than Mn-Mo-Ni low alloy steel. Ni and Cr additions increase the strength by solid solution hardening. In addition, microstructural changes from upper bainite to tempered martensite improve the strength of the low alloy steel by grain refining effect, and the changes in the precipitation behavior by Cr addition improve the ductile-brittle transition behavior along with a toughening effect of Ni addition.

• 전산구조공학
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• 제11권1호
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• pp.153-160
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• 1998

본 연구에서는 가압열충격 사고로 소형 냉각재 상실사고를 가정하여 냉각재의 온도와 압력의 이력으로 부터 용기 벽의 온도분포를 구하고, 이로 부터 열응력과 압응력을 해석적으로 구하였다. 또 균열 선단에서의 응력강도계수와 파괴인성치를 ASME코드의 방법을 이용하여 구하였고, 이들을 시간에 따라 비교하여 균열의 진전여부를 평가하였다. 원자로 용기 벽에 존재하는 여러 형태의 균열이 견딜 수 있는 최대 기준무연성천이온도를 결정하였으며 평가 결과에 대하여 고찰하였다.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.466-475
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• 2017

Coolant mixing under natural circulation flow regime constitutes a key parameter that may play a role in the course of an accidental transient in a nuclear pressurized water reactor. This issue has motivated some experimental investigations carried out within the OECD/NEA PKL projects. The aim was to assess the coolant mixing phenomenon in the reactor pressure vessel downcomer and the core lower plenum under several asymmetric steady and unsteady flow conditions, and to provide experimental data for code validations. Former studies addressed the mixing phenomenon using, on the one hand, one-dimensional computational approaches with cross flows that are not fully validated under transient conditions and, on the other hand, expensive computational fluid dynamic tools that are not always justified for large-scale macroscopic phenomena. In the current framework, an unsteady coolant mixing experiment carried out in the Rossendorf coolant mixing test facility is simulated using the three-dimensional porous media capabilities of the thermal-hydraulic system CATHARE code. The current study allows highlighting the current capabilities of these codes and their suitability for reproducing the main phenomena occurring during asymmetric transient natural circulation mixing conditions.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 1997년도 봄 학술발표회 논문집
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• pp.147-151
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• 1997

The dynamic characteristics of Control Element Drive Mechanism (CEDM) in Korea standard Nuclear Power Plant was reviewed as a secondary mass in a simplified two degree of freedom system, while the reactor vessel as a primary mass. The design improvement stratege to minimize each displacement amplitude of these primary and secondary masses was proposed. According to this stratege the designs of CEDM components, the shroud and the pressure housing, respectively, were changed using optimization technique.

• Nuclear Engineering and Technology
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• 제38권1호
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• pp.33-44
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• 2006

This paper presents a review of small-to-medium-sized, pressurized-water-cooled nuclear power reactors whose major primary coolant systems are integrated into a reactor pressure vessel, the concepts categorized as Integrated Primary System Nuclear Reactors (IPSRs). Typical examples of these proposals of interest in this review are CAREM, SMART, IRIS and IMR, all of which are being aimed at the near term deployment. Emphasis is placed on thermal hydraulic aspects. A brief characterization of the IPSR concepts is made and comparisons of plant key parameters are shown. Discussions will follow for the core cooling under rated power conditions and natural circulation heat removal on the basis of the design data available in the public domain.

• Nuclear Engineering and Technology
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• 제33권5호
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• pp.526-538
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• 2001

Fatigue tests in air and in room temperature water were performed to obtain comparable data and stable crack measuring conditions. In air environment, fatigue crack growth rate was increased with increasing temperature due to an increase in crack tip oxidation rate. In room temperature water, the fatigue crack growth rate was faster than in air and crack path varied on loading conditions. In simulated light water reactor (LWR) conditions, there was little environmental effect on the fatigue crack growth rate (FCGR) at low dissolved oxygen or at high loading frequency conditions. While the FCGR was enhanced at high oxygen condition, and the enhancement of crack growth rate increased as loading frequency decreased to a critical value. In fractography, environmentally assisted cracks, such as semi-cleavage and secondary intergranular crack, were found near sulfide inclusions only at high dissolved oxygen and low loading frequency condition. The high crack growth rate was related to environmentally assisted crack. These results indicated that environmentally assisted crack could be formed by the Electrochemical effect in specific loading condition.