• Nuclear Engineering and Technology
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• 제20권4호
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• pp.253-264
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• 1988

본 논문에서는 불란서에서 건설한 900 MWe급 가압경수형 원자로의 중성자 잡음해석 결과를 제시하였다. 중성자 잡음해석이란 노심내의 반응도 변화 및 노심의 수평운동으로 인한 노외검출기 신호의 변화를 해석하는 기법을 의미한다 이러한 방법은 Deterministic Dynamic Testing 기법중에서도 발전소의 정상운전 조건을 유지시키며 기존의 발전소 계측설비를 이용할 수 있다는 장점을 지니고 있다. 본 논문에 사용된 잡음신호는 울진 1호기 원자로의 시운전 시험기간에 구하였으며 이를 통계적 기술함수인 에너지 밀도함수(PSD), 검출기간의 상관함수 (CF)및 위상차(Phase Difference)로 나타내었다. 실험결과, 원자로 용기내의 냉각수 흐름 및 압력맥동 등에 의해 유도되는 Core Support Barrel(CSB)의 진동 주파수가 8Hz 근처임을 규명하였다.

• 한국전산구조공학회논문집
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• 제24권2호
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• pp.211-222
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• 2011

본 논문에서는 고준위폐기물 처분장에서 처분용기를 취급할 때 발생할 수 있는 운송차량에서 처분용기의 추락낙하 사고시 지면에 충돌할 때 지면으로부터 받는 충격력에 대하여 직경이 102cm인 가압경수로(PWR)용 처분용기에 대한 구조해석을 수행하여 처분용기의 구조적 안정성을 평가하였다. 이를 위하여 처분용기가 추락낙하하여 지면과 충돌 시에 처분용기가 받는 충격력을 구하기 위한 기구동역학해석을 상용 CAE 시스템인 RecurDyn을 이용하여 수행하였으며, 이와 같이 구한 충격력에 대하여 상용 유한요소해석 코드인 NISA를 이용하여 처분용기의 비선형구조해석을 수행하여 처분용기 내에 발생하는 응력 및 변형을 구하였다. 이를 바탕으로 처분용기가 처분장에서 취급 시 부주의로 운송차량에서 추락낙하 하는 경우 처분용기의 구조적 안전성을 평가하였다. 처분용기를 강체로 가정하고 기구동역학해석을 수행한 결과 처분용기는 지면과 두 가지 유형으로 충돌함을 알 수 있었고, 충돌 초기 지면으로부터 받는 충격력이 가장 크고 그 이 후 충돌 시에는 충격력이 점차로 감소함을 알 수 있었다. 안정적인 구조안전성 평가결과를 얻기 위하여 처분장에서 차량 운송 시 추락낙하 사고에서의 운송차량의 높이는 충분히 높은 5m로 가정하였다. 충격력에 대한 비선형구조해석은 추락낙하하여 가장 큰 값인 충돌 초기의 충격력의 크기를 가지고 비선형구조해석을 수행하였다. 해석결과 이송 중인 차량에서 추락낙하하는 경우 처분용기의 내부 주철삽입물에 주철의 항복응력보다 더 큰 응력이 발생하였으며, 이는 처분용기에 항복이 발생하여 경수로 처분용기의 구조적 안전성이 확보되지 못함을 보여주고 있다.

• Corrosion Science and Technology
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• 제20권4호
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• pp.189-195
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• 2021

A technology for designing and licensing a dedicated radiation shielding facility needs to be developed for safe and efficient operation an R&D center. Technology development is important for smooth operation of such facilities. Causes of damage to internal structures (such as baffle former bolt (BFB) of pressurized water reactor) of a nuclear power reactor should be analyzed along with prevention and countermeasures for similar cases of other plants. It is important to develop technologies that can comprehensively analyze various characteristics of internal structures of long term operated reactors. In high-temperature, high-pressure operating environment of nuclear power plants, cases of BFB cracks caused by irradiated assisted stress corrosion cracks (IASCC) have been reported overseas. The integrity of a reactor's internal structure has emerged as an important issue. Identifying the cause of the defect is requested by the Korean regulatory agency. It is also important to secure a foundation for testing technology to demonstrate the operating environment for medium-level irradiated testing materials. The demonstration testing facility can be used for research on material utilization of the plant, which might have highest fluence on the internal structure of a reactor globally.

• 설비공학논문집
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• 제29권12호
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• pp.654-662
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• 2017

When emergency core cooling system (ECCS) is operated during loss of coolant accident (LOCA) in a pressurized water reactor (PWR), pressurized thermal shock (PTS) phenomenon can occur as cooling water is injected into a cold leg, mixed with hot primary coolant, and then entrained into a reactor vessel. Insufficient flow mixing may cause temperature stratification and steam condensation. In addition, flow vibration may cause thermal stresses in surrounding structures. This will reduce the life of the reactor vessel. Due to the importance of PTS phenomenon, in this study, calculation was performed for Test 1 among six types of OECD/NEA ROSA tests with ANSYS CFX R.17. Predicted results were then compared to measured data. Additionally, because temperature difference between the hot coolant at the inlet of the cold leg and the cold cooling water at the inlet of the ECCS injection line is 200 K or more, buoyancy force due to density difference might have significant effect on thermal-hydraulic characteristics of flow. Therefore, in this study, the necessity to include buoyancy force term in governing equations for accurate prediction of single phase thermal stratification in both cold legs and downcomer by ECCS injection was numerically studied.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2002년도 제35회 춘계학술대회
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• pp.304-309
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• 2002

Fretting wear test in room temperature water was performed to evaluate the wear coefficient of Inconel 600,690 (Pressurized Water Reactor, PWR) and Alloy 800 (CANadian DeuteriumUranium, CANDU) steam generator (SG) tubes against ferritic and martensitic stainless steels. The main focus is to compare the wear behaviors between Alloy 800 and Inconel alloys. Test conditions are $10{\sim}30N$ of normal load, $200{\sim}450{\mu}m$ of sliding amplitude and 30Hz of frequency. The result indicated that the wear rate of Alloy 800 was higher than those of Inconel 690 at various test condition such as normal loads, sliding amplitudes etc. From the results of SEM observation, there was little evidence of plastic deformation layer that were dominantly formed on the worn surfaces of Inconel 690. Also, wear particles in Alloy 800 were released from contacting asperities deformed by severe plastic flow during fretting wear. Main cause of wear rate between Alloy 800 and Inconel 690 may be due to the difference of hardness between martensitic and ferritic stainless steel. The wear rate and wear mechanism of two tubes in room temperature water are discussed.

• Corrosion Science and Technology
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• 제11권4호
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• pp.141-150
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• 2012

Stress corrosion cracks in Alloy 600 compact tension specimens tested at $325^{\circ}C$ in a simulated primary water environment of pressurized water reactor were analyzed by analytical transmission electron microscopy and secondary ion mass spectroscopy (SIMS). From a fine-probe chemical analysis, oxygen was found on the grain boundary just ahead of the crack tip, and chromium oxides were precipitated on the crack tip and the grain boundary attacked by the oxygen diffusion, leaving a Cr/Fe depletion (or Ni enrichment) zone. The oxide layer inside the crack was revealed to consist of a double (inner and outer) layer. Chromium oxides existed in the inner layer, with NiO and (Ni,Cr) spinels in the outer layer. From the nano-SIMS analysis, oxygen was detected at the locations of intergranular chromium carbides ahead of the crack tip, which means that oxygen diffused into the grain boundary and oxidized the surfaces of the chromium carbides. The intergranular chromium carbide blunted the crack tip, thereby suppressing the crack propagation.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
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• pp.795-801
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• 1998

Metallic fatigue of Pressurized Water Reactor(PWR) materials is a generic safety issue for commercial nuclear power plants. It is very important to obtain the fatigue usage factor for component integrity and life extension. In this paper, fatigue usage was obtained at the inside surface of Kori unit 2, 3 and 4 RCP casing weld, based on the design transient. And it was intended to establish the procedure and the detailed method of fatigue evaluation in accordance with ASME Section III Code. According to this code rule, two methods to determine the stress cycle and the number of cycles could be applied. One method is the superposition of cycles of various design transients and the other is based on the assumption that a stress cycle correspond to only one design transient. Both method showed almost same fatigue usage in the RCP casing weld.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2001년도 봄 학술발표회 논문집
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• pp.259-266
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• 2001

This paper presents the results of a structural analysis to determine design variables such as the inner basket array type, and thicknesses of the outer shell and the lid and bottom of a spent nuclear fuel disposal canister. The canister construction type introduced here is a solid structure with a cast iron insert and a corrosion resistant overpack, which is designed for the spent nuclear fuel disposal in a deep repository in the crystalline bedrock, entailing an evenly distributed load of hydrostatic pressure from the groundwater and large swelling pressure from the bentonite buffer. Hence, the canister must be designed to withstand these large pressure loads. Many design variables may affect the structural strength of the canister. In this study, among those variables, the array type of inner baskets and thicknesses of outer shell and lid and bottom are attempted to be determined through a linear static structural analysis. Canister types studied here are one for the pressurized water reactor (PWR) fuel and another for the Canadian deuterium and uranium reactor (CANDU) fuel.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.463-478
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• 2019

The most important non-functional requirements for dependability of any Embedded Real-Time Safety Systems are safety, availability and reliability requirements. System architecture plays the primary role in achieving these requirements. Compliance with these non-functional requirements should be ensured early in the development cycle with appropriate considerations during architectural design. In this paper, we present an application of system architecture modeling for quantitative assessment of system dependability. We use probabilistic model checker (PRISM), for dependability analysis of the DTMC model derived from system architecture model. In general, the model checking techniques do not scale well for analyzing large systems, because of prohibitively large state space. It limits the use of model checking techniques in analyzing the systems of practical interest. We propose abstraction based compositional analysis methodology to circumvent this limitation. The effectiveness of the proposed methodology has been demonstrated using the case study involving the dependability analysis of safety system of a large Pressurized Water Reactor (PWR).

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.715-731
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• 2021

Uncertainties are calculated for pressurized water reactor (PWR) spent nuclear fuel (SNF) characteristics. The deterministic code STREAM is currently being used as an SNF analysis tool to obtain isotopic inventory, radioactivity, decay heat, neutron and gamma source strengths. The SNF analysis capability of STREAM was recently validated. However, the uncertainty analysis is yet to be conducted. To estimate the uncertainty due to nuclear data, STREAM is used to perturb nuclear cross section (XS) and resonance integral (RI) libraries produced by NJOY99. The perturbation of XS and RI involves the stochastic sampling of ENDF/B-VII.1 covariance data. To estimate the uncertainty due to modeling parameters (fuel design and irradiation history), surrogate models are built based on polynomial chaos expansion (PCE) and variance-based sensitivity indices (i.e., Sobol' indices) are employed to perform global sensitivity analysis (GSA). The calculation results indicate that uncertainty of SNF due to modeling parameters are also very important and as a result can contribute significantly to the difference of uncertainties due to nuclear data and modeling parameters. In addition, the surrogate model offers a computationally efficient approach with significantly reduced computation time, to accurately evaluate uncertainties of SNF integral characteristics.