• 한국산학기술학회논문지
• /
• 제14권7호
• /
• pp.3121-3126
• /
• 2013

원전 배관의 건전성평가는 원자로 안전을 위해 중요하며 결함발견 시 반드시 건전성을 확보해야만 한다. 균열을 갖는 구조물에 대한 정확한 응력확대계수 해석과 균열성장속도는 파괴강도와 피로수명을 평가하는데 필요로 한다. 피로설계와 수명평가는 배관, 산업공장장비 등과 같은 구조물을 설계하는데 극히 중요하다. 응력확대계수를 이용한 균열간의 상호 간섭해석은 유한요소법으로 구하였다. 내압을 받는 원통형구조물의 경우 표면균열의 인접점에서 간섭이 가장 크게 일어남을 확인하였다. 또한, 반복하중 균열에 대해서는 균열 성장평가와 더불어 피로하중에 의한 균열진전을 예측하기 위한 피로해석을 수행하였다.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2004년도 추계학술대회
• /
• pp.220-225
• /
• 2004

The mesh-insensitive structural stress procedure by Dong is modified to apply to the welded joints with local thickness variation and inignorable shear/normal stresses along local discontinuity surface. In order to make use of the structural stress based K solution for fatigue correlation of welded joints, a proper crack growth model needs to be developed. There exist some significant discrepancies in inferring the slope or crack growth exponent in the conventional Paris law regime. Two-stage crack growth model was not considered since its applications are focused upon the fatigue behavior in welded joints in which the load ratio effects are considered negligible. In this paper, a two-stage crack growth law considering high mean loading is proposed and proven to be effective in unifying the so-called anomalous short crack growth data.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2004년도 추계학술발표회 발표논문집
• /
• pp.1307-1308
• /
• 2004

• Corrosion Science and Technology
• /
• 제3권5호
• /
• pp.216-221
• /
• 2004

The operating experience showed that the fatigue is one of the major piping failure mechanisms in nuclear power plants (NPPs). The pressure and/or temperature loading transients, the vibration, and the mechanical cyclic loading during the plant operation may induce the fatigue failure in the nuclear piping. Recently, many fatigue piping failure occurred at the socket weld area have been widely reported. Many failure cases showed that the gap requirement between the pipe and fitting in the socket weld was not satisfied though the ASME Code Sec. III requires 1/16 inch gap in the socket weld. The ASME Code OM also limits the vibration level of the piping system, but some failure cases showed the limitation was not satisfied during the plant operation. In this paper, the fatigue behavior of the socket weld in the nuclear piping was estimated by using the three dimensional finite element method. The results are as follows. (1) The socket weld is susceptible to the vibration if the vibration levels exceed the requirement in the ASME Code OM. (2) The effect of the pressure or temperature transient load on the socket weld in NPPs is not significant because of the very low frequency of the transient during the plant lifetime operation. (3) 'No gap' is very risky to the socket weld integrity for the specific systems having the vibration condition to exceed the requirement in the ASME OM Code and/or the transient loading condition. (4) The reduction of the weld leg size from $1.09*t_1$ to $0.75*t_1$ can affect severely on the socket weld integrity.

• 대한기계학회논문집A
• /
• 제38권8호
• /
• pp.831-836
• /
• 2014

본 연구에서는 한국원자력연구원 내에 설치될 예정인 소듐시험 시설인 SELFA(Sodium Thermal-hydraulic Experiment Loop for Finned-tube Sodium-to-Air heat exchanger) 내에서 정상상태 가동온도가 $510^{\circ}C$의 고온 압력용기인 팽창탱크에 대해 고온 건전성 평가를 수행하였다. 팽창탱크에 대해 3 차원 유한요소해석에 기초하여 고온설계 기술기준인 ASME Section III Subsection NH 와 프랑스의 RCC-MRx 코드를 따라 크리프-피로 손상평가를 수행하였다. 평가결과 팽창탱크는 크리프-피로 설계 과도 하중 하에서 구조적 건전성을 유지하는 것으로 나타났다. 316L 스테인리스강 재질의 동 압력용기에 대해 정량적 코드 비교 분석을 수행하였다.

• 한국압력기기공학회 논문집
• /
• 제12권1호
• /
• pp.70-77
• /
• 2016

In this paper, the structural design and integrity evaluations for a reactor vessel of PGSFR sodium-cooled fast reactor(150MWe) are carried out in compliance with ASME BPV III, Division 5 Subsection HB. The reactor vessel is designed with a direct contact of primary sodium coolant to its inner surface and has a double vessel concept enclosing by containment vessel. To assure the structural integrity for 60 years design lifetime and elevated operating temperature of $545^{\circ}C$, which can invoke creep and creep-fatigue damage, the structural integrity evaluations are carried out in compliance with the ASME code rules. The design loads considered in this evaluations are primary loads and operation thermal cycling loads of normal heat-up and cool-down. From the evaluations, the PGSFR reactor vessel satisfies the ASME code limits but it was found that there is a little design margin of creep damage for inner surface at the region of cold pool free surface.

• 한국철도학회논문집
• /
• 제3권2호
• /
• pp.77-83
• /
• 2000

This study proposes the integrity evaluation of the bogie frame using ultrasonic fractography analysis. Analysis objectives in this study are to investigate fracture planes of damaged zone by the A-scan method. The surface condition of fracture planes shows degree of degradation by the stress concentration. The detection of the natural defects in the bogie frame is performed using the characteristics of echodynamic pattern in ultrasonic signal. Results of ultrasonic testing agree fairly well with those of actual fracture plane. In quantitative fractography analysis, microstructures of actual fracture plane turned out to be intergranular and transgranular fracture. Proposed ultrasonic fractography analysis in this study can be used for the integrity evaluation of the bogie frame.

• Journal of Mechanical Science and Technology
• /
• 제20권12호
• /
• pp.2061-2078
• /
• 2006

The main purpose of this paper is to establish the high temperature structural integrity evaluating procedures for the next generation reactors, which are to be operated at over 500$^{\circ}C$ and for 60 years. To do this, comparison studies of the high temperature structural design codes and assessment procedures such as the ASME-NH (USA), RCC-MR (France), DDS (Japan), and R5 (UK) are carried out in view of the accumulated inelastic strain and the creep-fatigue damage evaluations. Also the application procedures of the ASME-NH rules with the actual thermal and structural analysis results are described in detail. To overcome the complexity and the engineering costs arising from a real application of the ASME-NH rules by hand, all the procedures established in this study such as the time-dependent primary stress limits, total accumulated creep ratcheting strain limits, and the creep-fatigue damage limits are computerized and implemented into the SIE ASME-NH program. Using this program, the selected high temperature structures subjected to two cycle types are evaluated and the parametric studies for the effects of the time step size, primary load, number of cycles, normal temperature for the creep damage evaluations and the effects of the load history on the creep ratcheting strain calculations are investigated.

• Nuclear Engineering and Technology
• /
• 제53권7호
• /
• pp.2407-2417
• /
• 2021

A structural integrity evaluation program (STEP) was developed for the high temperature reactor design evaluation according to the ASME Boiler and Pressure Vessel Code (ASME B&PV), Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB. The program computerized HBB-3200 (the design by analysis procedures for primary stress intensities in high temperature services) and Appendix T (HBB-T) (the evaluation procedures for strain, creep and fatigue in high temperature services). For evaluation, the material properties and isochronous curves presented in Section II, Part D and HBB-T were computerized for the candidate materials for high temperature reactors. The program computerized the evaluation procedures and the constants for the weldment. The program can generate stress/temperature time histories of various loads and superimpose them for creep damage evaluation. The program increases the efficiency of high temperature reactor design and eliminates human errors due to hand calculations. Comparisons that verified the evaluation results that used the STEP and the direct calculations that used the Excel confirmed that the STEP can perform complex evaluations in an efficient and reliable way. In particular, fatigue and creep damage assessment results are provided to validate the operating conditions with multiple types of cycles.

• 한국기계가공학회지
• /
• 제13권5호
• /
• pp.107-115
• /
• 2014

Highly functional mill turrets have been developed and continuously improved to shorten the manufacturing time and enable multiple uses. Among these, a mill turret with B-axis rotary table was developed. The B-axis rotary table should be evaluated for structural integrity. Moreover, its worm and worm gear for transmitting power should be able to endure fatigue damage. Therefore, this article presents a structural analysis of this type of B-axis rotary table and confirms its static stability by comparing the stress results to the allowable stress levels. Next, the dynamic stability of the rotary table was investigated via a mode analysis and a harmonic analysis in a range determined by the results of a modal analysis. Finally, a worm gear set, the main part that drives the rotary table, is analyzed for fatigue and to estimate its lifetime. The results of the fatigue analysis allowed a prediction of the life of the worm gear set. The analytical results show that the B-axis rotary table has good structural integrity.