• 대한기계학회논문집A
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• 제26권1호
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• pp.23-30
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• 2002

Abstract The optimization is performed to reduce the mass of CNG pressure vessel reinforced with composite materials in the hoop direction. An axisymmetric shell element which takes into account the layered liner and hoop composite materials is thus developed and incorporated into a program Axicom. The accuracy of the program is then verified using the 4 noded element in ANSYS. Three different cases of optimization are then performed using the Axicom: (1) uniform hoop thickness, (2) varying hoop thickness, and (3) varying the ply angles and accordingly the thickness. Compared with a traditional method, cases (2) and (3) were found to be very effective in reducing the thickness and cost of the hoop composite materials by about 80% without sacrificing the safety factors.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.85-92
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• 2016

The operating temperature of VHTR components is much higher than that of conventional PWR due to high core outlet temperature of VHTR. Material requirements and technical issues of VHTR reactor components which are mainly dominated by high temperature service condition were discussed. The codification effort for high temperature material and design methodology are explained. The design class for VHTR components are classified as class A or B according to the recent ASME high temperature reactor design code. A separation of thermal boundary and pressure boundary is used for VHTR components as an elevated design solution. Key design characteristics for reactor pressure vessel, control rod, reactor internals, graphite reflector, circulator and intermediate heat exchanger were analysed. Thermo-mechanical analysis of the process heat exchanger, which was manufactured for test, is presented as an analysis example.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2002년도 학술대회지
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• pp.51-54
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• 2002

When a cold HPSI (High Pressure Safety Injection) fluid associated with an design basis accident, such as LOCA (Loss of Coolant Accident), enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena will arise due to incomplete mixing. If the stratified flow enters a reactor pressure vessel downcomer, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. Previous thermal-mixing analyses have assumed that the thermal stratification phenomena generated in stagnated loop of a partially stagnated coolant loop are neutralized in the vessel downcomer by strong flow from unstagnated loop. On the basis of these reasons, this paper presents the thermal-mixing analysis results in order to identify the fact that the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is affected by the strong flow of the unstagnated loop.

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

• 대한조선학회논문집
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• 제53권2호
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• pp.144-153
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• 2016

This paper aims to demonstrate the design, structure analysis, and hydrostatic pressure test of the cylinder used in 2000m water depth. The cylinder was designed in accordance with ASME pressure vessel design rule. The 1.5 times safety factor required by the general rule was applied to the design of the cylinder, because ASME rule is so excessive that it is not proper to apply to the hydrostatic pressure test. The finite element analysis was conducted for the cylinder. The cylinder was produced according to the design. The hydrostatic pressure test was conducted at the hyperbaric chamber in KRISO. The results of finite element analysis(FEM) and those of the hydrostatic pressure test were almost the same, which showed that the design was exact and reliable.

• 한국가스학회지
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• 제8권4호
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• pp.30-35
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• 2004

본 논문은 다구찌 실험 설계법을 이용하여 압력용기의 메탈시일 구조물에 대한 해석을 수행하였다. 다구찌 설계방법은 압력용기의 밀봉을 위한 캔티레버 타입 밀봉장치 구조물에 대한 설계 파라메터를 최적화 하는데 대단히 유용하다. 다구찌 기법으로 수행된 해석결과에 따르면, 최적화 설계 치수는 단지 16번의 반복 실험법에 의해 파라메터 치수를 최적의 조건으로 얻을 수 있다는 측면에서 간편한 설계도구이다. 이것은 다구찌 설계 실험법이 곡선면을 갖는 구조물의 최적화 설계에서 대단히 유용하다는 것을 의미한다. 다구찌 설계기법에 기초한 해석결과를 보면, 메탈시일 구조물의 에지부에 대한 최적화 치수와 경사각도는 $d_1=50mm,\;d_2=60mm,\;a_1=20^{\circ},\;a_2=8^{\circ},\;a_3=5^{\circ}$으로 각각 요약될 수 있다.

• 한국복합재료학회:학술대회논문집
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• 한국복합재료학회 2002년도 추계학술발표대회 논문집
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• pp.10-14
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• 2002

Structural behavior of high-pressure composite vessels of TYPE 3 (full-wrapped over a seamless aluminum liner) was studied through numerical simulations based on 3D nonlinear finite element method. Under high-pressure loading, a TYPE 3 composite vessel shows material nonlinearity due to elastic-plastic deformation of aluminum liner, and mismatch of deformation at the junction of cylinder and dome causes geometrical nonlinearity. Finite element modeling and analysis technique considering this nonlinearity was presented, and a pressure vessel of 6.8L of internal volume was analyzed. Design specification to satisfy requirements was determined based on analysis results.

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

• 한국압력기기공학회 논문집
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• 제10권1호
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• pp.107-112
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• 2014

The Code of Federal Regulations, Title 10, Part 50, Appendix H requires surveillance program for reactor pressure vessel(RPV) that the peak neutron fluence at the end of the design life of the vessel will exceed $1.0E+17n/cm^2$ (E>1.0MeV). 2D/1D Synthesis method based on DORT 3.1 transport calculation code has been widely used to determine fast neutron(E>1.0MeV) fluence exposure to RPV in the beltline region. RAPTOR-M3G(RApid Parallel Transport Of Radiation-Multiple 3D Geometries) performing full 3D transport calculation was developed by Westinghouse and KRIST(Korea Reactor Integrity Surveillance Technology) and applied for the evaluations of In-Vessel and Ex-Vessel neutron dosimetry. The reaction rates from measurement and calculation were compared and the results show good agreements each other.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2009년도 추계학술대회 논문집
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• pp.489-490
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• 2009