• Journal of Welding and Joining
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• v.1 no.2
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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.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.19 no.8
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• pp.44-54
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• 2018

Stress and fatigue of the distributor, an equipment of the high-pressure evaporator for the HRSG, were evaluated according to ASME Boiler & Pressure Vessel Code Section VIII Division 2. First, from the results of the piping system analysis model, reaction forces of the tubes connected to the distributor were derived and used as the nozzle load applied to the detailed analysis model of the distributor afterward. Next, the detailed model to analyze the distributor was constructed, the distributor being statically analyzed for the design condition with the steam pressure and the nozzle load. As a result, the maximum stress occurred at the bore of the horizontal nozzle, and the primary membrane stress at the shell and nozzle was found to be less than the allowable. Next, for the transient operating conditions given for the distributor, thermal analysis was performed and the structural analysis was carried out with the steam pressure, nozzle load, and thermal load. Under the transient conditions, the maximum stress occurred at the vertical downcomer nozzle, and of which fatigue life was evaluated. As a result, the cumulative usage factor was less than the allowable and hence the distributor was found to be safe from fatigue failure.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.1
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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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• v.54 no.3
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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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• v.55 no.12
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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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.4
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• pp.405-413
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• 2015

In the US, the number of cases of subterranean water contamination from tritium leaking through a damaged buried nuclear power plant pipe continues to increase, and the degradation of the buried metal piping is emerging as a major issue. A pipe blocked from corrosion and/or degradation can lead to loss of cooling capacity in safety-related piping resulting in critical issues related to the safety and integrity of nuclear power plant operation. The ASME Boiler and Pressure Vessel Codes Committee (BPVC) has recently approved Code Case N-755 that describes the requirements for the use of polyethylene (PE) pipe for the construction of Section III, Division 1 Class 3 buried piping systems for service water applications in nuclear power plants. This paper contains tensile and slow crack growth (SCG) test results for high-density polyethylene (HDPE) pipe welds under the environmental conditions of a nuclear power plant. Based on these tests, the fracture surface of the PENT specimen was analyzed, and the fracture mechanisms of each fracture area were determined. Finally, by using 3D finite element analysis, limit loads of HDPE related to premature failure were verified.