• 한국압력기기공학회 논문집
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• 제15권1호
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• pp.18-27
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• 2019

The Advanced Power Reactor 1400 (APR1400) Steam Generator (SG) uses alloy 690 as a tube material and SA-508 Grade 3 Class 1 as a tubesheet material to form tube-to-tubesheet joint through hydraulic expansion process. In this paper, the residual stresses in the SG tube-to-tubesheet contact area was investigated by applying Model-Based System Engineering (MBSE) methodology and the V-model. The use of MBSE transform system description into diagrams which clearly describe the logical interaction between functions hence minimizes the risk of ambiguity. A theoretical and Finite Element Methodology (FEM) was used to assess and compare the residual stresses in the tube-to-tubesheet contact area. Additionally, the axial strength of the tube to tubesheet joint based on the pull-out force against the contact joint force was evaluated and recommended optimum autofrettage pressure to minimize residual stresses in the transition zone given. A single U-tube hole and tubesheet with ligament thickness was taken as a single cylinder and plane strain condition was assumed. An iterative method was used in FEM simulation to find the limit autofrettage pressure at which pull-out force and contact force are of the same magnitude. The joint contact force was estimated to be 20 times more than the pull-out force and the limit autofrettage pressure was estimated to be 141.85MPa.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.1142-1153
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• 2019

This paper presents the effect of fusion procedure on the fusion performance of the thermal butt fusion in the safety class III buried HDPE piping per various tests performed, including high speed tensile impact, free bend, blunt notched tensile, notched creep, and PENT tests. The suitability of fusion joints and qualification procedures was evaluated by comparing test results from the base material and buttfusion joints. From the notched tensile test result, it was found that the fused joints have much lower toughness than the base material. It was also identified that the notched tensile test is more desirable than the high speed tensile impact and free bend tests presented in the ASME Code Case N-755-3 as a fusion qualification test method. In addition, with regard to the single low-pressure fusion joint performances, the procedure given by the ISO 21307 was determined to be better that the one specified in the Code Case N-755-3.

• 한국건설관리학회논문집
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• 제24권4호
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• pp.25-34
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• 2023

본 연구의 목적은 플랜트 건설에서 현장시공과 모듈시공에 대한 공사비를 산정하여 비교하는 것으로 공사비를 산정하는 대상을 Pipe Rack으로 한정하였다. 이에, 현장시공으로 준공된 국내 석유화학플랜트 건설사업 1곳을 사례로 선정한 후 비용자료를 조사하여 도출된 결과는 다음과 같다. Pipe Rack의 현장시공에 대한 직접공사비는 560억원으로 Steel Structure 251억원, Piping 308억원이며, 모듈시공에 대한 직접공사비는 607억원으로 Steel Structure 238억원, Piping 297억원으로 산정되었다. 또한, 현장시공과 모듈시공의 증감률을 비교해 보면, 재료비 1.9%, 경비 192.1% 증액되었으나, 노무비는 -9.1% 감액되어, 전체 직접공사비는 8.4%(47억원)가 증액되었다. 그리고 공사원가는 현장시공이 761억원, 모듈시공은 810억원으로 모듈시공이 6.4%(49억원) 증액되는 것으로 나타났다. 이와 같은 결과는 Pipe Rack을 모듈로 시공하는 경우 공사비가 증감되는 현황을 확인하기 위한 참고자료로 활용이 용이한 반면에, 모듈시공에 따른 간접적인 효과(노무인력 감소, 안전사고 발생 감소, 공사기간 단축 등)에 대한 연구가 필요하다.

• 한국압력기기공학회 논문집
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• 제9권1호
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• pp.70-76
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• 2013

Several finite element models for the leak-before-break (LBB) assessment of overlay dissimilar metal weldment were constructed and analyzed to develop a simple finite element modeling method. The J-integral, crack opening displacement (COD) and J-integral distribution along the crack front in thickness direction due to the applied moment were obtained from the analysis results of the constructed finite element models, and studied compared to the previous literatures. It is concluded that the modeling with base material only is simple and produces a slightly conservative results compared to the complex modeling composed with weld metal and base metal in the calculation of J-integrals and COD values which are used for the calculation of fracture toughness and postulated leakage crack length respectively.

• 한국압력기기공학회 논문집
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• 제5권2호
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• pp.7-12
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• 2009

The flaws detected by the non-destructive surface test methods shall be sized by means of the volumetric test such as an UT(ultrasonic test) or an ECT(eddy current test) for the purpose of analyzing and repairing them. It is generally known that the ECT is a comparatively effective technique for the small size cracks which are located shallowly from the surface. On this study, the ECT technique was tried to size the depth of the crack-like EDM notches, and it is identified that the ECT is an appropriate depth sizing technique for the shallow cracks less than 3mm in the Martensitic CA6NM material.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.131-136
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• 2003

CF8M cast austenitic stainless steel is used for several components such as primary coolant piping, elbow, pump casing, and valve bodies in light water reactors. These components are subject to thermal aging at the reactor operating temperature. Thermal aging results in spinodal decomposition of the delta-ferrite leading to increased strength and decreased toughness. In this study, three kinds of the aged CF8M specimen were prepared using an artificially simulated aging method. The objective of this study is to summarize the method of estimating ferrite contents, Charpy impact energy and J-R curve, and to evaluate the thermal embrittlement of the CF8M cast austenitic stainless steel piping used in the domestic nuclear power plants.

• 한국안전학회지
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• 제15권2호
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• pp.51-58
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• 2000

Recently, the J/T analysis based on elastic-plastic finite element analysis is popularly used in the nuclear industry to assess the integrity of a cracked pipe. The objective of this paper is to evaluate the effect of stress-strain curve for weld metal, variation of crack incremental length(${\delta}a$), and crack face pressure on the J/T analysis result. For this purpose, a parametric analysis was performed and the results calculated from finite element analysis were compared with those from the piping experimental data(stainless steel weldment pipe with circumferential through-wall crack). The numerical result using base metal material property is in agreement with the experimental one and the maximum load is decreased as the ${\delta}a$ for J/T analysis is increased.

• 한국압력기기공학회 논문집
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• 제8권2호
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• pp.1-6
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• 2012

A PHE (Process Heat Exchanger) in a nuclear hydrogen system is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature gas cooled Reactor) to the chemical reaction that yields a large quantity of hydrogen. A small-scale PHE prototype made of Hastelloy-X is being tested in a small-scale gas loop at Korea Atomic Energy Research Institute. Previous research on the high-temperature structural analysis of the small-scale PHE prototype had been performed only using parent material properties. In this study, high-temperature structural analysis using weld properties in weld zone was performed and the analysis results compared with the previous research.

• 한국압력기기공학회 논문집
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• 제4권2호
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• pp.1-6
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• 2008

Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source for nuclear hydrogen generation. The VHTR can produce hydrogen from heat and water by using a thermo-chemical process or from heat, water, and natural gas by steam reformer technology. A co-axial double-tube primary hot gas duct (HGD) is a key component connecting the reactor pressure vessel and the intermediate heat exchanger (IHX) for the VHTR. In this study, a preliminary design analysis for the primary HGD of the nuclear hydrogen system was carried out. These preliminary design activities include a determination of the size, a strength evaluation and an appropriate material selection. The determination of the size was undertaken based on various engineering concepts, such as a constant flow velocity model, a constant flow rate model, a constant hydraulic head model, and finally a heat balanced model.

• 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.