• 대한기계학회논문집
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• 제14권1호
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• pp.112-118
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• 1990

본 연구에서는 ASTM-E24.01.06에서 제안하고 있는 실험방법을 응용하여 균열 정지 파괴인성값을 측정하였다.즉 쐐기와 분리형 부싱(wedge and split bushing)으 로 압축하중을 가함으로 균열선 웨지하중 시편[crack line wedge loaded specimen(CL- WL시편)]에 인장력을 발생시켜서 균열정지 응력확대계수( $K_{1a}$)를 결정하였다. 그리고 균열개시 응력확대계수가 균열정지 응력확대계수에 미치는 영향들을 여러가지 재료들에 대하여 체계적으로 검토하였다.다.

• 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.

• Corrosion Science and Technology
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• 제20권5호
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• pp.308-314
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• 2021

Neutron dose level at bottom head of a reactor pressure vessel (RPV) was calculated using reactor vessel neutron transport for a Korean nuclear power plant A. At 34 EFPY with a 40-year (2042) design life after plating repair, irradiation fast neutron effect was 6.6x10¹⁵ n/cm². As helium(He) gas can be generated by Ni only at 1/10⁶ level of 5 × 10²¹ n/cm², He generation possibility in the Ni plating layer is very little during 40 years of operation (2042, 34 EFPY). Thermal neutrons can significantly affect the generation of He from Ni metal. At 10 years after a repair, He can be generated at a level of about 0.06 appm, a level that can add general welding repair without any consideration. After 40 years of repair, 9.8 appm of He may be generated. Although this is a rather high value, it is within the range of 0.1 to 10 appm when welding repair can be applied. Clad repair by Ni electroplating technology is expected to greatly improve the operation efficiency by improving the safety and shortening the maintenance period of the nuclear power plant.

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

• 비파괴검사학회지
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• 제19권6호
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• pp.433-438
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• 1999

원자로 압력용기 재료인 SA508 Steel을 온도 $70^{\circ}C$와 대기압하에서 최고 $10^{18}n/cm^2$까지 중성자를 조사시켜 조사량에 따른 미세경도 변화와 magneto-acoustic emission(MAE) 에너지를 측정하였다. 중성자 조사에 따른 경도의 변화는 조사량이 $10^{16}n/cm^2$까지는 거의 일정하였으나, 조사량이 $10^{17}n/cm^2$ 이상에서 급격히 증가하였다. MAE 에너지의 변화는 중성자 조사량에 따라 경도의 변화와 같은 형태로 변하였으나 그 변화량은 감소하여 그 변화의 추이는 경도의 변화와는 역의 형태였고, 또한 MAE 에너지의 상대적 변화와 경도 변화사이에는 아주 좋은 선형성을 보였다. 이러한 결과에서 SA508 강재는 $10^{17}n/cm^2$ 이상의 중성자에 조사될 경우에 재료에 중성자 조사에 의한 미세 결함이 급격히 증가하여 전위(dislocation)이동에 대한 저항성을 나타내는 마찰경화의 증가가 경도의 증가를 유발하고, 또한 이러한 미세 결함은 자기장과의 반응에서는 $90^{\circ}$ 자벽의 운동중에 자기탄성 변화를 유도하여 MAE 에너지의 감소를 유발함을 알 수 있었다. 그리고 경도의 변화량보다 MAE 에너지의 변화량이 더 크게 나타나, 중성자 조사에 의한 미세결함은 기계적 성질보다 자기적 성질에 더 민감하게 반응한다는 것을 알 수 있었다. 따라서 MAE가 중성자 조사에 의한 재료의 미세 구조 결함을 비파괴적인 방법으로 평가하는 강력한 도구의 가능성이 있음을 알 수 있었다.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1547-1554
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• 2017

VESTA (Verification of Ex-vessel corium STAbilization) and VESTA-S (-small) test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging $ZrO_2$ melt jet on a sacrificial material were performed to investigate the ablation characteristics. $ZrO_2$ melt in an amount of 65-70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40), and the other is a stainless steel (SUS304) melt. Metallic melt in an amount of 1.5-2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. $ZrO_2$ melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is $UO_2$ 60%, Zr 10%, $ZrO_2$ 15%, SUS304 14%, and $B_4C$ 1%, was melted in a cold crucible using an induction heating technique.

• 대한기계학회논문집A
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• 제26권3호
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• pp.445-451
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• 2002

The refractory-lined pipe is used to protect the system from high-temperature of the internal flow. The property of the refractory has an effect upon the stress analysis for fluid catalytic cracking(FCC) unit piping design. The equivalent elastic modulus and density considering steel and refractory must be applied in the stress analysis of the system. In the research, the theoretical method to obtain the value of the equivalent property is introduced and then the parametric analysis is carried out to understand the characteristic of the material properties, and the stress analysis is performed with reactor, the part of FCC unit.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.309-316
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• 2005

The contents of a new book currently in preparation are described. The dearth of books in the field of nuclear materials has left both students in nuclear materials classes and professionals in the same field without a resource for the broad fundamentals of this important sub-discipline of nuclear engineering. The new book is devoted entirely to materials problems in the core of light-water reactors, from the pressure vessel into the fuel. Key topics deal with the $UO_2$ fuel, Zircaloy cladding, stainless steel, and of course, water. The restriction to LWR materials does not mean a short monograph; the enormous quantity of experimental and theoretical work over the past 50 years on these materials presents a challenge of culling the most important features and explaining them in the simplest quantitative fashion. Moreover, LWRs will probably be the sole instrument of the return of nuclear energy in electric power production for the next decade or so. By that time, a new book will be needed.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2803-2811
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• 2020

With the aging of nuclear reactors, embrittlement of the reactor pressure vessel (RPV) steel, as a consequence of routine operations, is highly probable. To ensure operational integrity and safety, prediction and mitigation of compromising damage, brought on by pressurized thermal shock (PTS) following an emergency procedure, is of utmost importance. Computational fluid dynamics (CFD) codes can be employed to predict these events and have therefore been an acceptable method for such assessments. In this paper, CFD simulations of a density driven ECC state in the ROCOM facility are analyzed. Obtained numerical results are validated with the experimental measurements. Considerable attention is attributed to the boundary conditions and their influence, specifically outlet definitions, in order to determine and adequately replicate the non-active pumps in the facility. Consequent analyses focused on initial conditions as well as on the temporal discretization and inner iterations. Disparities due to different turbulent modelling approaches are investigated for standard RANS models. Based on observed trends for different cases, a definitive simulation setup has been established, results of which have been ultimately compared to the measurements.

• 한국압력기기공학회 논문집
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• 제7권2호
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• pp.42-47
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• 2011

The purpose of this study is to investigate the effects of through thickness on the mechanical properties and microstructural features in Mod. 9Cr-1Mo steels for RPVs. The microstructures at all locations were typically tempered martensite, but small amount of delta ferrite was observed at the center region. The prior austenite grain size increased with the depth from the surface. The yield strengths of center and 1/4T location were higher than that of surface by 30MPa. The impact toughness of center was low compared to those of other specimens. Also, upper shelf energy was low at the center. The toughness deterioration in center might be caused by larger size of the prior austenite grains and existence of the delta ferrite.