• 대한기계학회논문집A
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• 제40권5호
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• pp.429-435
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• 2016

ASS304 소재로 제작된 cold stretching 압력용기는 극저온 액화천연가스의 운반 및 저장에 이용된다. Cold stretching 압력용기는 특정 수압을 이용하여 허용응력까지 기존의 압력용기를 신장시킴으로써 재료에 상당량의 소성변형이 나타난다. 이 경우, ASS304 소재의 특성에 따라 소성변형 후 항복강도가 증가하며, 신장량에 따라 두께가 감소한다. 따라서 강도 및 무게측면에서 기존의 압력용기 대비 높은 효율을 나타낸다. 본 연구에서는 ASME 코드에 의거하여 제작된 소형 cold stretching 압력용기에서 직접 채취한 시편을 이용하여 극저온 인장 및 피로특성을 평가하였다. 또한 저온피로시험으로부터 획득한 S-N 선도를 통계적으로 접근하여 P-S-N 선도를 작성하였으며, SEM 을 이용하여 파단면을 분석하였다.

• 대한기계학회논문집A
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• 제41권9호
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• pp.809-815
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• 2017

압력용기 재료로 사용되는 STS 304강은 저온에서 기계적 특성이 우수하지만, 다른 합금과 비교하여 중량이 무겁다. 이러한 문제를 해결하는 방안으로 STS 304강에 콜드 스트레칭(Cold stretching)공법을 사용하는 방법이 있다. 본 연구에서는 ASME 규정에 따라 제작된 콜드 스트레칭 압력용기에서 직접 채취한 콜드 스트레칭부 및 용접부 시험편을 이용하여, 상온 및 $-170^{\circ}C$에서 응력비 0.1과 0.5에 대하여 컴플라이언스법을 이용하여 피로균열진전시험을 수행하였다. 시험결과, 용접부의 피로균열진전속도는 콜드 스트레칭부에 비해 동일 응력확대계수범위에서 빠른 것을 확인하였다. 이러한 결과는 콜드스트레칭 공법을 사용하여 제작한 STS강 압력용기 개발을 위한 기초자료로 활용될 수 있을 것이다.

• 비파괴검사학회지
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• 제33권4호
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• pp.368-375
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• 2013

소구경 배관 용접부 손상은 원자력발전소에서 빈번하게 발생되고 있는 문제이며, 이러한 손상은 보수를 위한 불시정지를 야기하고 결과적으로 발전소에 경제적인 손실을 초래한다. 따라서 소켓용접부의 제작결함 검출과 초기 성장 결함의 검출은 매우 중요하다고 할 수 있다. 지금까지 해당부위에 대한 검사는 ASME Code Section XI 요건에 의하여 표면검사를 적용하여 왔으나 강화검사로 체적검사기법인 초음파검사를 병행하여 수행하고 있다. 그러나 가동중에 발생되는 피로균열을 검출하기 위하여 적용되고 있는 일반 수동 초음파검사는 소켓용접부의 접근성과 초음파 탐촉자의 접촉 문제 등으로 인하여 검사 결과의 신뢰성에 문제가 있어 왔다. 본 연구에서는 위상배열 초음파검사 기술을 적용하여 소구경 배관 소켓용접부의 검사 신뢰성과 속도를 향상하고자 하였다. 이를 위하여 소구경 배관 소켓용접부 검사를 위한 3.5 MHz 위상배열 초음파 탐촉자를 제작하고 탐촉자의 접촉 조건과 일정한 신호 품질을 유지하기 위하여 수동 엔코더 스캐너를 개발하였다. 또한 검사 시스템을 구성하고 현장 검사를 위한 절차서도 개발하였다.

• 대한기계학회논문집A
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• 제21권7호
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• pp.1042-1049
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• 1997

LMFBR(Liquid Metal Fast Breeder Reactor) vessel is operated under the high temperatures of 500-550.deg. C. Thus, transient thermal loads were severe enough to cause inelastic deformation due to creep-fatigue and plasticity. For reduction of such inelastic deformations, Y-piece structure in the form of a thermal sleeve is used in LMFBR vessel under repeated start-up, service and shut-down conditions. Therefore, a systematic method for inelastic analysis is needed for design of the Y-piece structure subjected to such loading conditions. In the present investigation, finite element analysis of heat transfer and inelastic thermal stress were carried out for the Y-piece structure in LMFBR vessel under service conditions. For such analysis, ABAQUS program was employed based on the elasto-plastic and Chaboche viscoplastic constitutive equations. Based on numerical data obtained from the analysis, creep-fatigue damage estimation according to ASME Code Case N-47 was made and compared to each other. Finally, it was found out that the numerical predictio of damage level due to creep based on Chaboche unified viscoplastic constitutive equation was relatively better compared to elasto-plastic constitutive formulation.

• Journal of Welding and Joining
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• 제35권1호
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• pp.26-33
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• 2017

This study aims to investigate the suitability of requirement for post-weld heat treatment(PWHT) temperature when different P-No. materials are welded, which is defined by ASME Sec. III Code. For SA-516 Gr. 60 and SA-106 Gr. B carbon steels that are typical P-No. 1 material, simulated heat treatment were conducted for 8 h at $610^{\circ}C$, $650^{\circ}C$, $690^{\circ}C$, and $730^{\circ}C$, last two temperature falls in the temperature of PWHT for P-No. 5A low-alloy steels. Tensile and Charpy impact tests were performed for the heat-treated specimens, and then microstructure was analyzed by optical microscopy and scanning electron microscopy with energy-dispersive spectrometry. The Charpy impact properties deteriorated significantly mainly due to a large amount of cementite precipitation when the temperature of simulated heat treatment was $730^{\circ}C$. Therefore, when dissimilar metal welding is carried out for P-No. 1 carbon steel and different P-No. low alloy steel, the PWHT temperature should be carefully selected to avoid significant deterioration of impact properties for P-No. 1 carbon steel.

• Nuclear Engineering and Technology
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• 제34권3호
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• pp.187-201
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• 2002

The KN-12 cask is designed to transport 12 PWR spent nuclear fuels and to comply with the requirements of Korea Atomic Energy Act, IAEA Safety Standards Series No.57-1 and US 10 CFR Part 71 for a Type B(U)F package. It provides containment, radiation shielding, structural integrity, criticality control and heat removal for normal transport and hypothetical accident conditions. W.H 14$\times$14, 16$\times$16 and 17$\times$17 fuel assemblies with maximum allowable initial enrichment of 5.0 wt.%, maximum average burn-up of 50,000 MWD/MTU and minimum cooling time of 7 years being used in Korea will be loaded and subsequently transported under dry and wet conditions. A forged cylindrical cask body which constitutes the containment vessel is closed by a cask lid. Polyethylene rods for neutron shielding are arranged in two rows of longitudinal bore holes in the cask body wall. A fuel basket to accommodate up to 12 PWR fuel assemblies provides support of the fuels, control of criticality and a path to dissipate heat. Impact limiters to absorb the impact energy under the hypothetical accident conditions are attacked at the top and at the bottom side of the cask during transport. Handling weight loaded with water is 74.8 tons and transport weight loaded with water with the impact limiters is 84.3 tons. The cask will be licensed in accordance with Korea Atomic Energy Act 3nd fabricated in Korea in accordance with ASME B&PV Code Section 111, Division 3.

• 대한조선학회논문집
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• 제52권2호
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• pp.153-160
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• 2015

Pipelines are one of the most important structures in subsea equipment. It is the main equipment for transportation of crude oil and natural gas to the downstream facilities. Crude oil and natural gas leak will be carry out not only political and financial issues but also pollution to the environment. Inaccurate predictions of corrosion behavior will make hazardous consequences. The serviceability assessment of corroded structures is essential especially for subsea pipelines. As corrosion is concerned, the effects of failure due to significant reduction will make it hard to the pipeline operator to maintain the serviceability of pipelines. In this paper, the serviceability assessment of corroded crude oil pipeline is performed using the industry design code (Shell92, DNV RP F101, ASME B31G, BS 7910, PCORRC) and FEA depending on corrosion area. In last step, the future integrity of the subsea crude oil pipeline is assessed to predict the remaining year in service of crude oil pipelines.

• Nuclear Engineering and Technology
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• 제42권1호
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• pp.97-108
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• 2010

Fluid-elastic instability and turbulence-induced vibration of steam generator U-tubes of a nuclear power plant are studied numerically to investigate the effect of design changes of support structures in the upper region of the tubes. Two steam generator models, Model A and Model B, are considered in this study. The main design features of both models are identical except for the conditions of vertical and horizontal support bars. The location and number of vertical and horizontal support bars at the middle of the U-bend region in Model A differs from that of Model B. The stability ratio and the amplitude of turbulence-induced vibration are calculated by a computer program based on the ASME code. The mode shape with a large modal displacement at the upper region of the U-tube is the key parameter related to the fretting wear between the tube and its support structures, such as vertical, horizontal, and diagonal support bars. Therefore, the location and the number of vertical and horizontal support bars have a great influence on the fretting wear mechanism. The variation in the stability ratios for each vibrational mode is compared with respect to Model A and Model B. Even though both models satisfy the design criteria, Model A shows substantial improvements over Model B, particularly in terms of having greater amplitude margins in the turbulence-excited vibration (especially at the inner region of the tube bundle) and better stability ratios for the fluid-elastic instability.

• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1423-1432
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• 2016

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the "screening criterion" for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no "screening criterion". In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.

• 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.