• Water for future
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• v.23 no.4
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• pp.445-457
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• 1990

In order to predict the dispersion of a thermal discharge with strong turbulent and buoyant effects, the development of a numerical model using turbulence model and its application are significantly increased. In this study, a 3-dimensional steady-state model for the surface discharge of heated water into quiescent water body is developed. For the model closure of turbulent terms the 4-equation turbulence model is used. For economic numerical simulation, the elliptic governing equations are transformed to the partially parabolic equations. In general, the simulated results by the present model agree well to the experimental results by Pande and Rajaratnam. The model characteristics are presented in comparison with the predicted results of the 2-equation turbulence model by McGuirk and Rodi. Applying the 4-equation turbulence model to the Korea nuclear unit 1 at Kori site, feasibility and efficiency of the present model are validated.

• Computational Structural Engineering
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• v.10 no.4
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• pp.155-164
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• 1997

A reactor pressure vessel, which contains fuel assemblies and reactor vessel internals, has the thermal stress resulting from the cool-down and heat-up of the vessel wall in combination with the pressure stress from system pressure resulting in large stresses. The combination of the pressure stress and thermal stress along with a decrease in fracture toughness may cause through-wall propagation of a relatively small crack. Therefore, it is necessary to define the relations between operating pressure and temperature during cool-down and heat-up. In this study, theory of fracture mechanics for a pressure/temperature limit curve is investigated and a numerical procedure for generating it is developed. Plant-specific limit curves for the Kori unit 1 plant, the oldest nuclear power plant in Korea, have been obtained for several cooling and heating rates and their results are discussed.

• Nuclear Engineering and Technology
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• v.17 no.4
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• pp.279-289
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• 1985

The purpose of this paper is to investigate the density-wave oscillation type instability in the recirculating loop of U-tube steam generator (UTSG). The perturbed and nodalized conservations equations based on the drift-flux model have been derived to obtain the single-and two-phase pressure drop perturbations, by taking into account the slip between phases, nonuniform heat flux and heated wall dynamics. To assess the stability, the frequency domain technique with the Nyquist criterion has been used under the constant pressure drop boundary condition through the loop. The computer implementation of this model, SASG, was used for the parametric study of the steam generator in Kori-Unit 1. The results of the parametric study revealed important factors influencing UTSG stability margin.

• Nuclear Engineering and Technology
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• v.18 no.2
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• pp.71-77
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• 1986

A one dimensional neutron kinetics program, BIK which is applicable to the safety analyses of PWR's is developed to analyze the reactor core in axial dimension. The BIK employs the finite difference technique in space and $\theta$-time integration method in time. Detailed models for the Doppler and moderator feedbacks and control rod motion are included. The benchmark of the nuclear model is carried out through the ANL benchmark problem and the time dependent nuclear power change in the rod ejection accident of KNU1 is calculated by BIK code. The results indicate that the BIK can predict the neutron dynamics with fair accuracy within the limits of one dimensional analysis and it is useful for the safety analyses of PWR's.

• Journal of Radiation Protection and Research
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• v.40 no.3
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• pp.162-167
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• 2015

Although the perfomance indicators of the nuclear power plants in Korea show optimal, it requires detailed analysis and discussion centered on the radiation dose. As analysis methods, analysis on the radiation dose of nuclear power plants over the past five years was assessed by comparing the relevant radiation dose of radiation workers and per capita average annual radiation dose of the world's major nuclear power stations was also analyzed. The radiation workers over the annual radiation dose limit of 50 mSv were not. The contrast ratio of the radiation exposure according to the reactor type was the normal operation of PHWR was 6.2% higher than those of the PWR. This shows the radiation work of PHWR during normal driving operation is much more than those of PWR. According to the Performance Indicators of the World Association of Nuclear Operator, the annual radiation dose per unit in 2013 showed 527 man-mSv of Korea is the best country among the major nuclear power generating states, the world average was 725 man-mSv. The annual per capita radiation dose is about 80% less than 1 mSv of the public dose limit and also the average per capita dose showed a very low level as 0.82 mSv. Workers in related organizations showed 1.07 mSv, the non-destructive inspection agency workers showed 3.87 mSv. The remarkable results were due to radiation reduced program such as development of radiation shielding and radiation protection. In conclusion, the radiation exposured dose of nuclear power plants workers in Korea showed a trend which is ideally reduced. But more are expected to be difficul and the psychological insecurity against the operation of the nuclear power plants is existed to the residents near the nuclear power plants. So the radiation dose reduction policy and radiation dose follow up study of nuclear power plants will be continously excuted.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.381-387
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• 2003

Hull arising from the spent PWR fuel elements is classified as a high-level radioactive waste. This report describes the radio-chemical characteristics of the hull -from PWR spent fuel of 32, 000MWd/tU burn-up and 15 years cooling, discharged from Gori Unit I cycled 4 -7-by examination and literature survey. On the basis of the results, a method of degradation to middle and low-level radio active waste was proposed by dry process such as laser or plasma technique with removing the nuclides deposited on the surface of the hull.

• Proceedings of the Korea Technology Innovation Society Conference
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• 2017.11a
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• pp.1473-1487
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• 2017

2017년 6월 19일 고리 원전 1호기 영구정지 기념식에서 문재인 대통령의 '탈원전' 선언을 계기로, 국내에서는 후쿠시마 원전 사고 이후 탈원전으로 복귀한 독일의 사례에 대한 관심이 고조되었다. 독일은 1986년 체르노빌 원전 사고 이후 탈원전에 대한 논의가 본격적으로 이루지면서 집권 정부의 성향과 사회 경제적 요인에 따라 "2000년 탈원전 선언 ${\rightarrow}$ 2010년 탈원전 보류 ${\rightarrow}$ 2011년 탈원전 복귀"의 결정 과정을 거쳤다. 이러한 정책 변화의 배경에는 간헐성의 재생에너지를 뒷받침(back-up)할 수 있는 자국의 풍부한 갈탄 매장량, 지리적으로 주변국과 연결된 전력망을 통해 전력을 상시 주고받을 수 있는 전력 수급 환경, 탈원전에 대한 정부 국민 산업계의 40여년에 걸친 합의형성 등 '독일 자국의 실정을 반영한 정책적 판단'이 자리하고 있다. 그럼에도 불구하고, 2011년 후쿠시마 원전 사고 직후 독일의 즉각적인 탈원전 복귀는 화석연료 사용의 증가로 인한 온실가스 배출량 증가, 재생에너지 보조금 증가 및 송전망 확대로 인한 전기요금 상승, 간헐성의 재생에너지로 인한 불안정한 전력 수급, 과잉 생산된 전력의 수출로 인한 주변국 전력계통 혼란 등의 문제를 초래하고 있다. 이에 본고에서는 독일의 탈원전 정책이 '어떤 정책결정 과정을 거쳤으며, 현실적으로 어떤 문제에 직면해 있는지'를 살펴보고 우리나라의 에너지 수급 현실을 반영한 정책적 시사점을 도출하고자 한다.

• Nuclear Engineering and Technology
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• v.19 no.1
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• pp.34-41
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• 1987

This work concerns with a comparison of the 1-dimensional (or 1-D) load follow analysis method with reference to the detailed 3-dimensional (or 3-D) computations. For this purpose a 1-D two-group finite difference code, HLOFO, and a 3-D coarse-mesh code based on the modified Borresen's method, CMSNAC, are developed. The CMSNAC code is used to obtain the 3-D power peaks and reactivity parameters in response to power swing from 100 to 50 and back to 100% in the 12-3-6-3 load cycle for the BOL of the KORI Unit 1 PWR core. The 3-D result is then compared with the 1-D HLOFO computations, the cross section and buckling inputs of which are obtained by combining the flux-volume weighting scheme with the approximate flux from the auxiliary 3-D computations. It is shown that the 1-D computation has a limited accuracy, yet it is confirmed that it can describe the core axial average behavior which is fairly consistent with the detailed 3-D computation.

• Nuclear Engineering and Technology
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• v.9 no.4
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• pp.223-236
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• 1977

An integral computer code has been developed for a mechanical and thermal design and performance analysis of an oxide fuel rod in a pressurized water reactor. The code designated as FROD 1.0 takes into account the phenomena of radial power depression within the pellet, cracking, densification and swelling of the pellet, fission gas release, clad creep, pellet-clad contact, heat transfer to coolant and buildup of corrosion layers on the clad surface. The FROD 1.0 code yields two-dimensional temperature distributions, dimensional changes, stresses, and internal pressure of a fuel rod as a function of irradiation time within a reasonable computation time. The code may also be used for the analyses of oxide fuel rods in other thermal reactors. As an application of FROD 1.0 the behavior of fuel rod loaded in the first core of Go-ri Nuclear Power Plant Unit 1 is predicted for the two power histories corresponding to steady state operation and Codition II of the ANS Classification. The results are compared with the design criteria described in the Final Safety Analysis Report and a discrepancy between these two values is discussed herein.

• Nuclear Engineering and Technology
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• v.13 no.1
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• pp.1-11
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• 1981

The KINS (KAERI-Improved Nodal Simulation) program, a three-dimensional nodal simulation code for pressurized water reactors, has been developed and benchmarked against the first cycle of the Kori-1 reactor. The KINS program is based on the computational model used in FLARE code and has been modified to represent the PWR characteristics more explicitly. The critical boron concentration and three-dimensional power distribution at the beginning of life hot zero power have been calculated and compared with the operating data. A three-dimensional depletion calculation at the intervals of 1000 MWD/MTU turnup steps has been performed. As the result of comparison, our calculation is shown to be in excellent agreement with the operating data. It is displayed that, incorporated with the computing time, the KINS program is an effective and powerful tool for PWR core management.