• 한국정밀공학회:학술대회논문집
• /
• 한국정밀공학회 2010년도 추계학술대회 논문집
• /
• pp.715-716
• /
• 2010

• Nuclear Engineering and Technology
• /
• 제44권6호
• /
• pp.683-688
• /
• 2012

The structural integrity of mechanical components during several transients should be assured in the design stage. This requires a fatigue analysis including thermal and stress analyses. As an example, this study performs a fatigue analysis of the reactor pressure vessel of SMART during arbitrary transients. Using heat transfer coefficients determined based on the operating environments, a transient thermal analysis is performed and the results are applied to a finite element model along with the pressure to calculate the stresses. The total stress intensity range and cumulative fatigue usage factor are investigated to determine the adequacy of the design.

• 한국추진공학회지
• /
• 제7권2호
• /
• pp.36-43
• /
• 2003

탄소섬유 T700/Epoxy로 만든 직교이방성 구조로 된 두께가 얇은 압력용기의 반경방향의 변위에 관한 해를 곡선좌표계의 평형방정식을 사용하여 구하였다. 3차원 곡선좌표계의 변형율과 변위의 관계를 간단히 하면서 지배방정식을 유도하기 위해 변분이론과 가상일의 원리를 사용하였다. 다른 여러 종류의 직교이방성 압력용기에 대한 계산 결과를 제시했으며 수압시험을 한 결과와 비교 검토하였다. 계산결과와 시험결과는 비교적 잘 일치하였다.

• Journal of Welding and Joining
• /
• 제18권2호
• /
• pp.100-104
• /
• 2000

Pressure vessels usually consist of main body and pipes which are connected with the main body. And as joining method of such main body and pipes, welding is carried out. After welding, welding residual stresses inevitably occur around welded joints. As residual stresses act harmfully on fatigue strength, corrosion and buckling strength of structure, PWHT is carried out for the purpose of removing the residual stress. But, during PWHT process, $2\frac{1}{4}Cr-1Mo$ steels are frequently apt to generate reheat crack. For this reason, it is strongly needed to analyze and examine the mechanical behavior of welded joints before and after PWHT process. So, in this study, welded nozzle parts of pressure vessel where reheat cracks frequently occur are selected for examining the mechanism of crack-occurrence.

• Nuclear Engineering and Technology
• /
• 제55권11호
• /
• pp.4146-4158
• /
• 2023

This paper proposes a combined plastic and creep constitutive model of A533B1 pressure vessel steel to simulate progressive deformation of nuclear pressure vessels under severe accident conditions. To develop the model, recent tensile test data covering a wide range of temperatures (from RT to 1,100 ℃) and strain rates (from 0.001%/s to 1.0%/s) was used. Comparison with experimental data confirms that the proposed combined plastic and creep model can well reflect effects of temperature and strain rate on tensile behaviour up to failure. In the companion paper (Part II), the proposed model will be used to simulate OECD lower head failure (OLHF) test data.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
• /
• pp.310-315
• /
• 1997

System-Integrated Modular Advanced Reactor (SMART), where major primary components such as modular helical steam generator and self regulating pressurizer are integrated into reactor vessel, is currently under development. The pressurizer is designed to control the primary pressure mainly with partial pressure of nitrogen gas and to maintain the fluid temperature as low as possible for the purpose of minimizing steam contribution. The steam generator (SG) is designed to produce super-heated steam inside tube at power operation. Because the in-vessel pressurizer and in-vessel SG are classified as the characteristic components of SMART, it is important to perform a steady state calculation of these components in order to evaluate the adoption of these components. A steady state analysis of the in-vessel pressurizer and in-vessel SG has been performed under normal power operation and the results show an acceptable performance of the components.

• 한국정밀공학회지
• /
• 제16권7호
• /
• pp.24-31
• /
• 1999

Bulging is a forming method to shape of die cavity by using hydraulic pressure in tube or vessel. Bulging machine and die were developed in order to produce vessel for keeping warm. Bulging machine is a double type with two horizontal cylinders for bulging of two pieces at the same time. The developed die system has one bulging die and two drawing dies for necking at the both ends of tube. The diameter of tube expands by hydraulic pressure in tube. at the same time, thrust at the both ends of tube. pushes tube in the direction of expansion to obtain high expanding rate with no crack. In this study, the bulging properties were investigated to solve tube crack and necking in manufacturing vessel by the combination method of bulging and drawing. As a result, high expanding rate of tube radius without crack, precision necking and high productivity were obtained.

• International Journal of Precision Engineering and Manufacturing
• /
• 제2권4호
• /
• pp.40-46
• /
• 2001

Bulging is a forming method to shape die cavity by using hydraulic pressure in tube or vessel. Bulging machine and die were developed in order to produce vessel for keeping warm. Bulging machine is a double type with two horizontal cylinders for bulging of two pieces at the same time. The developed die system has one bulging die and two drawing dies for necking at both ends of the tube. The diameter of tube expands by hydraulic pressure in tube. At the same time, thrust at both ends of the tube pushes tube in the direction of expansion to obtain high expansion rate with no crack. In this study, the bulging properties were investigated to solve tube crack and necking in manufacturing vessel by combining bulging and drawing. As a result, high expanding rate of tube radius without crack, precision necking and high productivity were obtained.

• 한국압력기기공학회 논문집
• /
• 제15권2호
• /
• pp.40-49
• /
• 2019

In this paper, failure probabilities of the OPR1000 reactor vessel under pressurized thermal shock (PTS) were estimated using the probabilistic fracture mechanics code, R-PIE. Input variables of initial crack distribution, crack size, copper contents, and upper shelf toughness were selected for the sensitivity analyses. A wide range of the input data were considered. Through-wall cracking frequencies determined by the product of the vessel failure probability and the corresponding occurrence frequency of the transient were also compared to the acceptance criterion. The results showed that transient history had the most significant impact on the vessel failure probability. Moreover, conservative assumptions resulted in extremely high through-wall cracking frequencies.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2004년도 추계학술대회
• /
• pp.383-387
• /
• 2004

The Fuel Test Loop (FTL) consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). The test condition in IPS such as pressure, temperature and quality of the main cooling water, can be controlled by the OPS. The FTL has been developed to be able to irradiate three pins to the core irradiation hole (IR1 hole) by considering for its utility and user's irradiation requirement. The IPS vessel assembly (IVA) consists of IPS head, outer pressure vessel, inner pressure vessel, inner assembly and test fuel carrier. The IVA is approximately 5.6 m long and fits within a 74 mm in diameter envelope over the full height of the chimney. Above the top of the chimney, the head of the IPS is enlarged to allow the closure flanges and pipe work connections. IVA was designed to test the CANDU and PWR nuclear fuel pin together. Specially, wished to minimize interference by nuclear fuel change in design and synthesize these items and shape design for IVA.