• Nuclear Engineering and Technology
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• 제19권1호
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• pp.22-33
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• 1987

원자력 발전소에 있어서 정상가동 상태나 이상동작시에 핵연료 피복관의 건전성 확보와 연관하여 피복재의 항복거동은 중요한 문제이다. 급격한 출력상승 상황에서 이산화 우라늄 소결체와 피복관 사이의 노내 조사거동의 차이는 소결체와 피복관 사이에 Contact Pressure를 야기 시킨다. 만일 이 Contact Pressure가 Zircaloy 피복관의 Yield Pressure에 도달하면 피복관에는 영구변형이 일어난다. 이 변형은 원자로의 출력이 정상상태로 회복되더라도 존재하므로 소결체와 피복관 사이의 Gap을 증대시킨다. 이러한 상황을 묘사하기 위해 본 논문에서는 구리 Mandrel과 Zircaloy사이의 열팽창 차이를 이용하는 Mandrel 팽창 실험을 실행했다. 실험 결과 측정된 Zircaloy 피복관의 외경 팽창치와 본 논문에서 유도된 수학적 관계식들을 이용하여 온도에 따른 Zircaloy 피복관의 내부항복압력과 항복응력, 피복재의 항복에 따른 핵연료 소결체와 피복관 사이의 Gap 증대를 구하고, 항복 거동에 따른 온도의 영향을 보기 위해 항복과정의 활성화 에너지를 구했다. 본 실험과 분석에서 얻어진 이들 결과들은 다른 실험 결과들과 상당히 일치하였으며, 이것으로 볼 때 본 논문에서 유도된 관계식들과 Mandrel 팽창 실험이 Zircaloy 피복관의 항복거동과 Gap Expansion 측정에 신뢰성이 있음을 알 수 있었다.

• Nuclear Engineering and Technology
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• 제41권2호
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• pp.171-178
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• 2009

The rate of delayed hydride cracking (DHC), V, has been measured in cold-worked and stress-relieved Zircaloy-4 fuel cladding using the Pin-Loading Tension technique. At $250^{\circ}C$ the mean value of V from 69 specimens was $3.3({\pm}0.8)x10^{-8}$ m/s while the temperature dependence up to $275^{\circ}C$ was described by Aexp(-Q/RT), where Q is 48.3 kJ/mol. No cracking or cracking at very low rates was observed at higher temperatures. The fracture surface consisted of flat fracture with no striations. The results are compared with previous results on fuel cladding and pressure tubes.

• Nuclear Engineering and Technology
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• 제26권3호
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• pp.389-400
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• 1994

FLECHT SEASET 실험 데이터를 사용하여 대형 냉각재 상실 사고시 재관수 첨두 피복재 온도에 대한 RELAP5 /MOD3/KAERI의 예측능력을 평가하였으며, 관련 불확실성을 통계적으로 정량화 하였다. 중력구동 재관수 실험및 광범위한 재관수율, 시스템 압력, 초기 피복재 온도, 연료봉 출력을 포괄하는 강제구동 재관수 실험들로 구성된 18개의 실험이 평가에 사용되었다. 평가 결과 재관수 첨두 피복재 온도에 대해 평균 7.56 K 낮게 예측하였으며 이를 포함한 관련 불확실성의 상한은 95% 신뢰도 수준에서 약 99 K로 정량화 되었다.

• 대한기계학회논문집A
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• 제25권7호
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• pp.1139-1146
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• 2001

The reactor pressure vessel(RPV) is usually cladded with stainless steel to prevent corrosion and radiation embrittlement, and a number of subclad cracks have been found during an in-service-inspection. These subclad cracks should be assured for a safe operation under normal conditions and faulted conditions such as pressurized thermal shock(PTS). Currently available integrity assessment procedure for an RPV, ASME Code Sec. XI, are built on the basis of linear fracture mechanics (LEFM). In PTS condition, however, thermal stress and mechanical stress give rise to high tensile stress at the cladding and elastic-plastic behavior is expected in this area. Therfore, ASME Code Sec. XI is overly conservative in assessing the structural integrity under PTS condition. In this paper, the fracture parameter (stress intensity factor, K, and RT(sub)NDT) from elastic analysis using ASME Sec. XI and finite element method were validated against 3-D elastic-plastic finite element analyses. The difference between elastic and elastic-plastic analysis became significant with increasing crack depth. Therfore, it is recommended to perform elastic-plastic analysis for the accurate assessment of subclad cracks under TPS which causes plastic deformation at the cladding.

• Wind and Structures
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• 제30권1호
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• pp.99-117
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• 2020

This paper focuses on the processes of wind flow in atmospheric boundary layer, to produce realistic full scale pressures for design of low-rise buildings. CFD with LES turbulence closure is implemented on a scale 1:1 prototype building. A proximity study was executed computationally in CFD with LES that suggests new recommendations on the computational domain size, in front of a building model, apart from common RANS-based guidelines (e.g., COST and AIJ). Our findings suggest a location of the test building, different from existing guidelines, and the inflow boundary proximity influences pressure correlation and reproduction of peak loads. The CFD LES results are compared to corresponding pressures from open jet, full scale, wind tunnel, and the ASCE 7-10 standard for roof Component & Cladding design. The CFD LES shows its adequacy to produce peak pressures/loads on buildings, in agreement with field pressures, due to its capabilities of reproducing the spectral contents of the inflow at 1:1 scale.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2638-2651
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• 2020

Nozzle corner cracks present at the intersection of reactor pressure vessels (RPVs) and inlet or outlet nozzles have been a persistent problem for a number of years. The fracture analysis of such nozzle corner cracks is very important and critical for the efficient design and assessment of the structural integrity of RPVs. This paper aims to perform an engineering critical assessment of RPVs with nozzle corner cracks subjected to several transients accompanied by pressurized thermal shocks. The critical crack size of the RPV model with nozzle corner cracks under transient loading is evaluated on failure assessment curve. In particular, the influence of cladding on the crack initiation of nozzle corner crack under thermal transients is studied. The influence of primary internal pressure and secondary thermal stress on the stress field at nozzle corner and SIF at crack front is analyzed. Finally, the influence of different crack size and crack shape on the final critical crack size is analyzed.

• 한국압력기기공학회 논문집
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• 제19권2호
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• pp.163-170
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• 2023

Since spent nuclear fuel assemblies (SFA) are transported to interim storage or final disposal facility after cooling the decay heat, finite element analysis (FEA) with simplification is widely used to show their integrity against cladding failure to cause dispersal of radioactive material. However, there is a lack of research addressing the comprehensive impact of shape and element simplification on analysis results. In this study, for the optimization of a typical pressurized water reactor SFA, different types of finite element models were generated by changing number of fuel rods, fuel rod element type and assembly length. A series of FEA in use of these different models were conducted under a shock load data obtained from surrogate fuel assembly transportation test. Effects of number of fuel rods, element type and length of assembly were also analyzed, which shows that the element type of fuel rod mainly affected on cladding strain. Finally, an optimal finite element model was determined for other practical application in the future.

• Tribology and Lubricants
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• 제23권3호
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• pp.130-133
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• 2007

Fuel cladding tubes in nuclear fuel assembly are held up by supporting grids because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube and support. The fretting wear of tube and support can threaten the safety of nuclear power plant. Therefore, a research about the fretting wear characteristics of tube-support is required. The fretting wear tests were performed with supporting grids and cladding tubes, especially after corrosion treatment on tubes, in water. The tests were done using various applied loads with fixed amplitude. From the results of fretting tests, the wear amounts of tube materials can be predictable by obtaining the wear coefficient using the work rate model. Due to stick phenomena the wear depth was changed as increasing load and temperature. The maximum wear depth was decreased as increasing the water temperatures. At high temperatures there are the regions of some severe adhesion due to stick phenomena.

• Tribology and Lubricants
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• 제24권3호
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• pp.129-132
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• 2008

Fuel cladding tubes in nuclear fuel assembly are held up by supporting grids because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube and support. The fretting wear of tube and support can threaten the safety of nuclear power plant. Therefore, a research about the fretting wear characteristics of tube-support is required. The fretting wear tests were performed with supporting grids and cladding tubes, especially after corrosion treatment on tubes, in water. The tests were done using various applied loads with fixed amplitude. From the results of fretting tests, the wear amounts of tube materials can be predictable by obtaining the wear coefficient using the work rate model. Due to stick phenomena the wear depth was changed as increasing load and temperature. The maximum wear depth was decreased as increasing the water temperatures. At high temperatures there are the regions of some severe adhesion due to stick phenomena.

• Nuclear Engineering and Technology
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• 제35권4호
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• pp.356-368
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• 2003

Material properties such as coolant specific heat, film heat transfer coefficient, cladding thermal conductivity, surface diffusion coefficient of the multi-bubble are improved in MACSIS-Mod1. The axial power and flux profile module was also incorporated with irradiation history. The performance and feasibility of the updated driver fuel pin have been analyzed for nominal parameters based on the conceptual design for the KALIMER breakeven core by MACSIS-MOD1 code. The fuel slug centerline temperature takes the maximum at 700mm from the bottom of the slug in spite of the nearly symmetric axial power distribution. The cladding mid-wall and coolant temperatures take the maximum at the top of the pin. Temperature of the fuel slug surface over the entire irradiation life is much lower than the fuel-clad eutectic reaction temperature. The fission gas release of the driver fuel pin at the end of life is predicted to be $68.61\%$ and plenum pressure is too low to cause cladding yielding. The probability that the fuel pin would fail is estimated to be much less than that allowed in the design criteria. The maximum radial deformation of the fuel pin is $1.93\%$, satisfying the preliminary design criterion ($3\%$) for fuel pin deformation. Therefore the conceptual design parameters of the driver fuel pin for the KALIMER breakeven core are expected to satisfy the preliminary criteria on temperature, fluence limit, deformation limit etc.