• Nuclear Engineering and Technology
• /
• 제23권1호
• /
• pp.20-29
• /
• 1991

본 연구에서는 2030년까지의 전력수요, 전력생산중 원자력발전의 비중, 기존 원전표준화 계획, 국내제작 능력을 반영하여 개량형 경수로와 중수로 (CANDU)에 대한 참조 시나리오를 도출하고 각 참조 시나리오와 핵연료주기 전략별 핵연류주기 비용, 원자력 발전 단가, 우라늄 소요량, 인력 소요량을 계산하였다. 참조 시나리오들에 대한 분석을 한 결과 우라늄 자원활용, 원전안전성, 인력활용 측면이 노형 전략수립의 주요 인자로 작용하며 발전단가는 전략별로 큰 차이가 없는 것으로 나타났다.

• Nuclear Engineering and Technology
• /
• 제52권2호
• /
• pp.417-428
• /
• 2020

Investigations of the commercial aircraft impact effect on nuclear island infrastructures have been drawing extensive attention, and this paper aims to perform the safety assessment of Generation III nuclear power plant (NPP) buildings subjected to typical commercial aircrafts crash. At present Part III, the local damage of the rigid components of aircraft, e.g., engine and landing gear, impacting the steel concrete (SC) structures of NPP containment is mainly discussed. Two typical SC target panels with the thicknesses of 40 mm and 100 mm, as well as the steel cylindrical projectile with a mass of 2.15 kg and a diameter of 80 mm are fabricated. By using a large-caliber air gas gun, both the projectile penetration and perforation test are conducted, in which the striking velocities were ranged from 96 m/s to 157 m/s. The bulging velocity and the maximal deflection of rear steel plate, as well as penetration depth of projectile are derived, and the local deformation and failure modes of SC panels are assessed experimentally. Then, the commercial finite element program LS-DYNA is utilized to perform the numerical simulations, by comparisons with the experimental and simulated projectile impact process and SC panel damage, the numerical algorithm, constitutive models and the corresponding parameters are verified. The present work can provide helpful references for the evaluation of the local impact resistance of NPP buildings against the aircraft engine.

• Nuclear Engineering and Technology
• /
• 제53권7호
• /
• pp.2184-2194
• /
• 2021

The Free-Piston Stirling engine (FPSE) is of interest for many research in aerospace due to its advantages of long operating life, higher efficiency, and zero maintenance. In this study, a 1-kW FPSE was proposed by analyzing the requirements of Space Reactor Power Systems (SRPS), of which performance was evaluated by developing a code through the Simple Analysis Method. The results of SAM showed that the critical parameters of FPSE could satisfy the designed requirements. The heater of the FPSE was designed with the copper rectangular fins to enhance heat transfer, and the parametric study of the heater was performed with Computational Fluid Dynamics (CFD) software STAR-CCM⁺. The Performance Evaluation Criteria (PEC) was used to evaluate the heat transfer enhancement of the fins in the heater. The numerical results of the CFD program showed that pressure drop and Nusselt number ratio had a linear growth with the height of fins, and PEC number decreased as the height of fins increased, and the optimum height of the fin was set as 4 mm according to the minimum heat exchange surface area. This paper can provide theoretical supports for the design and numerical analysis of an FPSE for SRPSs.

• 방사성폐기물학회지
• /
• 제9권3호
• /
• pp.149-159
• /
• 2011

원전 운영자는 계통 및 기기의 열화 등에 의해서 발생할 수 있는 비계획적 방사성물질의 방출로 인한 환경 및 주변 주민에의 영향을 합리적으로 달성 가능한 한 낮게 유지하기 위해서 비계획적 방출을 조기에 감지할 수 있는 부지에 적합한 지하수 감시 프로그램을 수립해야 하며, 이를 위해서는 해당 부지의 수문지질 특성의 파악을 통해 지하수 유동을 평가해야 한다. 본 논문에서는 고리 1호기에서 계획되고 있는 지하수 감시 프로그램에 필요한 자료를 제공하기 위해, 고리 1발전소 부지의 기존 수문지질 조사 및 관련 자료의 조사를 통해 부지 내 지하수 유동특성을 파악하고 이를 바탕으로 가상의 비계획적 방출에 의한 삼중수소의 오염운(汚染雲)의 거동을 모의하였다. 모의 결과 지하수의 주 유동 방향은 남서방향이었으며 지하수의 대부분이 남쪽 및 동쪽 바다로 유입되었다. 삼중수소 오염운 역시 바다로 향하였으나 지하집수조(dewatering sump)에 의해 그 속도가 지연되는 것을 확인할 수 있었다.

• 한국지진공학회논문집
• /
• 제20권7_spc호
• /
• pp.443-451
• /
• 2016

A Gyeongju earthquake in the magnitude of 5.8 on the Richter scale (the moment magnitude of 5.4), which was recorded as the strongest earthquake in Korea, occurred in September 12, 2016. Compared with the 2011 Virginia earthquake, the moment magnitude was slightly smaller and its duration was 3 seconds, much shorter than 10 seconds of the Virginia earthquake, resulting in relatively minor damage. But the two earthquakes are quite similar in terms of the overall scale, unexpectedness, and social situation. The North Anna Nuclear Power Plant, which is a nuclear power plant located at 18 km away from the epicenter of the Virginia earthquake, had no damage to nuclear reactors because the reactors were automatically shut down as the design basis earthquake value was exceeded. Ground accelerations of the 2016 Gyeongju earthquake did not exceed the threshold value but the manual shutdown was carried out so that Wolsong Nuclear Power Site was not damaged. Damaged historic homestead house and masonry structures due to the Virginia earthquake have been repaired, reinforced, and rebuilt based on a long-term earthquake recovery project. Likewise, it will be necessary to carefully carry out an earthquake recovery planning program to improve overall seismic performance and to reconstruct the historic buildings and structures damaged as a result of the Gyeongju earthquake.

• Nuclear Engineering and Technology
• /
• 제50권1호
• /
• pp.88-96
• /
• 2018

Concerns over reliability assessments of the main components in nuclear power plants (NPPs) related to aging and continuous operation have increased. The conventional reliability assessment for main components uses experimental correlations under general conditions. Most NPPs have been operating in Korea for a long time, and it is predictable that NPPs operating for the same number of years would show varying extent of aging and degradation. The conventional reliability assessment does not adequately reflect the characteristics of an individual plant. Therefore, the reliability of individual components and an individual plant was estimated according to operating data and conditions. It is essential to reflect aging as a characteristic of individual NPPs, and this is performed through prognostics. To handle this difficulty, in this paper, the general path model/Bayes, a data-based prognostic method, was used to update the reliability estimated from the generic database. As a case study, the authors consider the aging for steam generator tubes in NPPs and demonstrate the suggested methodology with data obtained from the probabilistic algorithm for the steam generator tube assessment program.

• Journal of Nuclear Fuel Cycle and Waste Technology
• /
• 제1권1호
• /
• pp.1-7
• /
• 2013

The safe management of radioactive waste is a national task required for sustainable generation of nuclear power and for energy self-reliance in Korea. Since the initial introduction of nuclear power to Korea in 1978, rapid growth in nuclear power has been achieved. This large nuclear power generation program has produced a significant amount of radioactive waste, both low- and intermediate-level waste (LILW) and spent nuclear fuel (SNF); and the amount of waste is steadily growing. For the management of LILW, the Wolsong LILW Disposal Center, which has a final waste disposal capacity of 800,000 drums, is under construction, and is expected to be completed by June 2014. Korean policy about how to manage the SNF has not yet been decided. In 2004, the Atomic Energy Commission decided that a national policy for SNF management should be established considering both technological development and public consensus. Currently, SNF is being stored at reactor sites under the responsibility of plant operator. The at-reactor SNF storage capacity will run out starting in 2024. In this paper, the fundamental principles and steps for implementation of a Korean policy for national radioactive waste management are introduced. Korean practices and prospects regarding radioactive waste management are also summarized, with a focus on strategy for policy-making on SNF management.

• Nuclear Engineering and Technology
• /
• 제40권5호
• /
• pp.349-364
• /
• 2008

Over the years, probabilistic risk assessment (PRA) research activities conducted at the U.S. Nuclear Regulatory Commission (NRC) have played an essential role in support of the agency's move towards risk-informed regulation. These research activities have provided the technical basis for NRC's regulatory activities in key areas; provided PRA methods, tools, and data enabling the agency to meet future challenges; supported the implementation of NRC's 1995 PRA Policy Statement by assessing key sources of risk; and supported the development of necessary technical and human resources supporting NRC's risk-informed activities. PRA research aimed at improving the NRC's understanding of risk can positively affect the agency's regulatory activities, as evidenced by three case studies involving research on fire PRA, human reliability analysis (HRA), and pressurized thermal shock (PTS) PRA. These case studies also show that such research can take a considerable amount of time, and that the incorporation of research results into regulatory practice can take even longer. The need for sustained effort and appropriate lead time is an important consideration in the development of a PRA research program aimed at helping the agency address key sources of risk for current and potential future facilities.

• Nuclear Engineering and Technology
• /
• 제48권2호
• /
• pp.339-350
• /
• 2016

Based on the resonance integral (RI) tables produced by the NJOY program, the conventional subgroup method usually ignores both the resonance elastic scattering and the resonance interference effects. In this paper, on one hand, to correct the resonance elastic scattering effect, RI tables are regenerated by using the Monte Carlo code, OpenMC, which employs the Doppler broadening rejection correction method for the resonance elastic scattering. On the other hand, a fast resonance interference factor method is proposed to efficiently handle the resonance interference effect. Encouraging conclusions have been indicated by the numerical results. (1) For a hot full power pressurized water reactor fuel pin-cell, an error of about +200 percent mille could be introduced by neglecting the resonance elastic scattering effect. By contrast, the approach employed in this paper can eliminate the error. (2) The fast resonance interference factor method possesses higher precision and higher efficiency than the conventional Bondarenko iteration method. Correspondingly, if the fast resonance interference factor method proposed in this paper is employed, the $k_{inf}$ can be improved by ~100 percent mille with a speedup of about 4.56.

• Nuclear Engineering and Technology
• /
• 제53권8호
• /
• pp.2547-2555
• /
• 2021

The frequency of reactor coolant leakage is expected to increase over the lifetime of a nuclear power plant owing to degradation mechanisms, such as flow-acceleration corrosion and stress corrosion cracking. When loss of coolant accidents (LOCAs) occur, several parameters change rapidly depending on the size and location of the cracks. In this study, leak flow during LOCAs is predicted using a deep fuzzy neural network (DFNN) model. The DFNN model is based on fuzzy neural network (FNN) modules and has a structure where the FNN modules are sequentially connected. Because the DFNN model is based on the FNN modules, the performance factors are the number of FNN modules and the parameters of the FNN module. These parameters are determined by a least-squares method combined with a genetic algorithm; the number of FNN modules is determined automatically by cross checking a fitness function using the verification dataset output to prevent an overfitting problem. To acquire the data of LOCAs, an optimized power reactor-1000 was simulated using a modular accident analysis program code. The predicted results of the DFNN model are found to be superior to those predicted in previous works. The leak flow prediction results obtained in this study will be useful to check the core integrity in nuclear power plant during LOCAs. This information is also expected to reduce the workload of the operators.