• Journal of Welding and Joining
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• 제1권2호
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• pp.45-52
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• 1983

Many pressure vessels for the hot H$\sub$2//H$\sub$2/S service are made of 2+1/4Cr-1Mo steel with austenitic stainless steel overlay to combat agressive corrosion due to hydrogen sulfide. Hydrogen dissolves in to materials during operation, and sometimes gives rise to unfore-seeable damages. Appropriate precautions must, therefore, be taken to avoid the hydrogen induced damages in the design, fabrication and operation stage of such reactor vessels. Recently, hydrogeninduced cracking (or Disbonding) was found at the interface between base metal and stainless weld overlay of a desulfurizing reactor. Since the stainless steel overlay weld metal is subjected to thermal and internal-pressure loads in reactor operation, it is desirable for the overlay weld metal to have high strength and ductility from the stand point of structural safety. In section III of ASME Boiler and Pressure Vessel Code, Post-Weld Heat Treatment(PWHT) of more than one hour per inch at over 1100.deg. F(593.deg. C) is required for the weld joints of low alloy pressure vessel steels. This heat treatment to relieve stresses in the welded joint during construction of the pressure vessel is considered to cause sensitization of the overlay weld metal. The present study was carried out to make clear the diffusion of carbon migration by PWHT in dissimilar metal welded joint. The main conclusion reached from this study are as follows: 1) The theoretical analysis for diffusion of carbon in stainless steel overlay weld metal does not agree with Fick's 2nd law but the general law of molecular diffusion phenomenon by thermodynamic chemical potential. 2) In the stainless steel overlay welded joint, the PWHT at 720.deg. C for 10 hours causes a diffusion of carbon atoms from ferritic steel into austenitic steel according to the theoretical analysis for carbon migration and its experiment. 3) In case of PWHT at 720.deg. C for 10 hours, the micro-hardness of stainless steel weld metal in bonded zone increase very highly in the carburized layer with remarkable hardening than that of weld metal.

• 한국산학기술학회논문지
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• 제19권8호
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• pp.44-54
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• 2018

복합발전플랜트 배열회수보일러 고압증발기의 기기인 분배기에 대하여 설계조건과 과도운전조건을 고려하여 응력 및 피로에 관한 안전성을 평가하였다. 먼저, 배열회수보일러 튜브군 모델의 해석결과로부터 분배기의 상부에 연결되는 수직 강수관, 하부에 연결되는 수직 급수배관, 열교환기의 입구헤더로 향하는 수평방향의 방사형 배관들에 대하여 노즐하중을 도출하였다. 이와 같이 구한 노즐하중은 분배기의 상세모델에 대한 설계조건과 과도운전조건의 해석 시에 노즐 단면에 가해지는 하중으로 사용하였다. 분배기의 상세한 해석모델을 만들고 설계조건의 내압과 노즐하중에 대한 정적구조해석을 수행하였다. 설계조건에서 최대응력은 수평방향 배관의 노즐 보어에서 발생하였다. 최대응력 위치의 국부 1차 막응력이 쉘과 노즐에서 허용기준보다 작으므로 ASME Code의 허용기준을 만족하는 것으로 나타났다. 배열회수보일러에 주어진 8가지 과도운전조건을 고려하여, 분배기의 상세모델에 대하여 열해석을 수행하고, 과도운전 시의 내압, 노즐하중, 열하중에 대한 과도구조해석을 수행하였다. 과도운전조건에서 최대응력은 분배기 상부의 수직 강수관 노즐 부위에서 발생하였다. ASME Code에 의거하여 수직 강수관 노즐 부위의 피로수명을 평가하였다. 결과적으로 계산된 누적피로사용계수가 허용기준보다 작으므로 기대수명 동안에 피로파손에 관하여 안전한 것으로 나타났다.

• 대한기계학회논문집A
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• 제26권1호
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• pp.179-187
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• 2002

Finite element analysis was performed to evaluate the integrity of the tubes in the fixed tubesheet of horizontal type heat exchanger under operating condition. For the finite element analysis of the heat exchanger, tubes and tubesheets were equivalently modeled with concentroidal hexagonal columns and solid plates having equivalent properties for the convenience of finite element modeling, respectively. Load combination of tube pressure and thermal expansion most likely to precipitate possible failure of the tubes was selected and applied to the finite element analysis. The compressive stresses of the tubes were calculated based on displacements of each tube, which were obtained from anile element analysis. Finally, the maximum tube stress was compared with the design criterion of ASME Boiler and Pressure Vessel Code Section VIII.

• 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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• 제55권12호
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• pp.4477-4490
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• 2023

Fuel locking mechanisms (FLMs) are essential in upward-flow research reactors to prevent accidental fuel separation from the core during reactor operation. This study presents a novel design concept for a remotely controlled plate-type nuclear fuel locking mechanism. By employing electromagnetic field analysis, we optimized the design of the electromagnet for fuel unlocking, allowing the FLM to adapt to various research reactor core designs, minimizing installation space, and reducing maintenance efforts. Computational flow analysis quantified the drag acting on the fuel assembly caused by coolant upflow. Subsequently, we performed finite element analysis and evaluated the structural integrity of the FLM based on the ASME boiler and pressure vessel (B&PV) code, considering design loads such as dead weight and flow drag. Our findings confirm that the new FLM design provides sufficient margins to withstand the specified loads. We fabricated a prototype comprising the driving part, a simplified moving part, and a dummy fuel assembly. Through basic operational tests on the assembled components, we verified that the manufactured products meet the performance requirements. This remote-controlled micro locking mechanism holds promise in enhancing the safety and efficiency of plate-type nuclear fuel operation in upflow research reactors.

• 대한기계학회논문집A
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• 제39권4호
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• pp.405-413
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• 2015

최근 미국에서는 가동기간이 오래된 원전 매설배관에서 부식 및 침식에 의해 삼중수소 누설로 지하수가 오염되는 사례가 급증하고 있다. 따라서, 현재 원전 안전등급 매설배관으로 사용되고 있는 금속재료의 배관을 대신해서 부식 및 침식 등의 열화 손상에 대한 저항성이 우수한 고밀도 폴리에틸렌(HDPE) 배관을 ASME Code Class 3 안전계통 배관으로 사용하기 위한 연구가 수행되고 있다. 본 연구에서는 발전소 가동 중 매설배관에 가해질 수 있는 하중과 온도 범위를 바탕으로 HDPE 배관 융착부에 대한 인장 시험과 저속균열성장 (SCG) 시험을 수행하였다. 시험 결과로 얻은 SCG 시험편의 파단면을 분석하여 HDPE 재료의 파손 기구를 파악하였다. 이를 바탕으로 3D 유한요소 해석을 이용하여 균열이 있는 HDPE 재료가 버틸 수 있는 한계하중에 대한 검증을 수행하였다.