• Korean Journal of Plant Resources
• /
• v.20 no.1
• /
• pp.63-68
• /
• 2007

The purpose of this study was to applicate fundamental data for breeding new cultivar through selection of superior individuals and to investigate its nut characteristics in walnut hybrids. Selection of superior individuals with good nut qualities including high nut weight (NWT, >13g) and percentage of weight relative to total weight of nut (PWR, >50%) was carried out and then 4 promising individuals ($Sansung 4{\times}Concord-8,\;Concord{\times}Sansungl,\;Concord{\times}Sansung4\;and\;McKinster{\times}Punghan1$) were selected by quantitative characters. Especially width of pad of suture was main factor in selection of nut with high PWR (percentage of weight relative to total weight of nut) In addition, kernel length showed high correlation with kernel and nut weight. Thickness of septem (TOP) also had an effect on ease of kernel removal (EKR). In qualitative characters, there is a great difference among the individuals and cross combinations as well as showed simultaneously various characteristics in the same individual.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.6 no.2
• /
• pp.147-154
• /
• 2008

It is expected that the temporary storage facilities at the nuclear power plants will be full of the spent fuels within 10 years. Provided that a centralized interim storage facility is constructed along the coast of the Korean peninsula to solve this problem, a substantial amount of spent fuels should be transported by sea or by land every year. In this paper we developed a computer program for the analysis of transportation logistics of the spent fuels from 4 different nuclear power plant sites to the hypothetical centralized interim storage facility and the final repository. Mass balance equations were used to analyze the logistics between the nuclear power plants and the interim storage facility. To this end a computer program, CASK, was developed by using the VISUAL BASIC language. The annual transportation rates of spent fuels from the four nuclear power plant sites were determined by using the CASK program. The parameter study with the program illustrated the easiness of logistics analysis. The program could be used for the cost analysis of the spent fuel transportation as well.

• Nuclear Engineering and Technology
• /
• v.42 no.1
• /
• pp.79-88
• /
• 2010

Spent $UO_2$ nuclear fuel discharged from a nuclear power plant (NPP) contains fission products, U, Pu, and other actinides. Due to neutron capture by $^{238}U$ in the rim region and a temperature gradient between the center and the rim of a fuel pellet, a considerable increase in the concentration of fission products, Pu, and other actinides are expected in the pellet periphery of high burnup fuel. The characterization of the radial profiles of the various isotopic concentrations is our main concern. For an analysis, spent nuclear fuels originating from the Yeonggwang-2 pressurized water reactor (PWR) were chosen as the test specimens. In this work, the distributions of some actinide isotopes were measured from center to rim of the spent fuel specimens by a radiation shielded laser ablation inductively coupled plasma mass spectrometer (LA-ICP-MS) system. Sampling was performed along the diameter of the specimen by reducing the sampling intervals from 500 ${\mu}m$ in the center to 100 ${\mu}m$ in the pellet periphery region. It was observed that the isotopic concentration ratios for minor actinides in the center of the specimen remain almost constant and increase near the pellet periphery due to the rim effect apart from the $^{236}U$ to $^{235}U$ ratio, which remains approximately constant. In addition, the distributions of local burnup were derived from the measured isotope ratios by applying the relationship between burnup and isotopic ratio for plutonium and minor actinides calculated by the ORIGEN2 code.

• Journal of Radiation Protection and Research
• /
• v.11 no.2
• /
• pp.91-103
• /
• 1986

Experimental study for the leaching behavior of Cs-137 was carried out using the simulated evaporator bottom product of PWR plant. The method of leach test proposed by the IAEA was partially modified using ANS method. The effect of various factors, i.e., sampling method, curing temperature, curing time, leachant temperature, vermiculite addition and volume-to-surface ratio, was considered in this experiment. Diffusion model in semi-infinite slab was in a good agreement with the data obtained from 4-weeks cured specimens. The effective diffusion coefficient of the specimens which were cured at the temperature of $24^{\circ}C$ for 4 weeks was found to be $1.20{\sim}1.47{\times}10^{-11}cm^2/sec$. With the experimentally obtained diffusion coefficient ($1.47{\times}10^{-11}cm^2/sec$), long-term prediction for the leaching of Cs-137 was carried out using finite-slab approximation. The estimated fraction of Cs-137 which remains in the environment is found to be less than 0.25 percent of initial amount after 100 years. About 25 years after the beginning of leaching, its fractional amount in the environment reachs the maximum value, 0.66 percent of initial amount.

• Nuclear Engineering and Technology
• /
• v.8 no.2
• /
• pp.101-116
• /
• 1976

An attempt is made to analyze the unit nuclear power cost of the Go-ri units 1 and 2 in terms of a set of model data. For the calculational purpose, the power cost is first decomposed into the cost components related to the plant capital, operation and maintenance, working capital requirements, and fuel cycle operation. Then, POWERCO-50 computer code is applied to enumerate the first three components and MITCOST-II is used to evaluate the fuel cycle cost component. The specific numerical results are the fuel cycle cost of Go-ri unit 2 for three alternative fuel cycles presumed, levelized unit power cost of units 1 and 2, and the sensitivity of the power cost to the fluctuation of the model data. Upon comparision of the results with the power cost of the fossil power plants in Korea, it is found that the nuclear power is economically preferred to the fossil power. Nevertheless, the turnkey contract value of Go-ri unit 2 appears to be rather expensive compared with the available data on the construction cost of the PWR plants. Therefore, it is suggested that, in order to make the nuclear power plants more attractive in Korea, the unfavorable contract of such kind must be avoided in the future introduction of the nuclear power plant. Capacity factor is of prime importance to achieving the economic generation of the nuclear electricity from the Go-ri plant. Therefore, it is concluded that more efforts should be directed to make the maximum use of the Go-ri plant.

• Proceedings of the KSME Conference
• /
• 2001.11a
• /
• pp.16-21
• /
• 2001

The PTS reference temperature of reactor pressure vessel for one of the Korean NPPs has been predicted to exceed the screening criteria before it reaches it's design life. To cope with this issue, a plant-specific PTS analysis had been performed in accordance with the Regulatory Guide 1.154 in 1999. As a result of that analysis, it was found that current methodology of RG. 1.154 was very conservative. The objective of this study is to examine the effects of changing various input parameters and to determine the amount of conservatism of the current PTS analysis method. To do this, based on the past PTS analysis experience, parametric study were performed for various models using modified VISA-II code. This paper discusses the analysis results and recommendations to reduce the conservatism of current analysis method.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05b
• /
• pp.795-801
• /
• 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.

• Proceedings of the KSME Conference
• /
• 2001.06a
• /
• pp.316-321
• /
• 2001

A previous feasibility study for the Korean lead plant, PLiM Phase I, showed a strong possibility of continued operation beyond the original licensed period. In 1998, PLiM Phase II study was initiated aimed at performing additional detailed evaluations on a wider range of components. The objective of this paper is to present the Korean PLiM efforts for Class 1 piping which is identified as one of the critical components with regard to long-term operation. The key findings such as typical design features, degradation mechanisms, technical issues, draft results from the lifetime evaluation for Class 1 piping of the lead plant are briefly described.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
• /
• 2003.05a
• /
• pp.473-483
• /
• 2003

본 논문은 안전해석 등에 사용되는 RETRAN-3D 등 최적해석 코드를 기반으로 하면서도 복잡한 하드웨어 없이 간편한 GUI (Graphic User Interface)를 이용하여 광범위한 발전소 과도상태를 해석하기 위한 다양한 기능을 통해 시뮬레이션 조작을 쉽게 할 수 있는 웨스팅하우스형 950MW급 최적 원전운전분석기 (Nuclear Plant Analyzer)를 다루고자 한다. WH형 950MW 원전 운전최적분석기는 기존의 단순한 Point Kinetics 모델이 아닌 정교한 3D 실시간 노심모델과 RETRAN 코드를 기반으로 하는 실시간 NSSS 열수력 모델 (ARTS)이 통합된 모델을 갖추고 있으며, 해당형식발전소 (WH 3 Loop PWR Plant : 고리 3,4호기, 영광1,2호기 원전)의 여러 가지 과도사고를 실시간으로 정상, 비정상, 비상운전 등으로 모의할 수 있도록 개발되었다. 모의결과 주요 과도 상태의 결과가 해석한 결과와 잘 일치하였으며, 해당형식 발전소 과도 분석이나 규제요원 훈련에 이용될 계획이다.

• Proceedings of the KIEE Conference
• /
• 1993.07a
• /
• pp.261-264
• /
• 1993

This paper proposes an adaptive generalized predictive control with feedforward algorithm for steam generator level control in nuclear power plant. The proposed algorithm is shown that the parameters of N-step ahead predictors can be obtained using the parameters of one-step ahead predictor which is derived from plant model with feedforward. Using this property the proposed scheme is an adaptive algorithm which consists of GPC method and the recursive least squares algorithm for identifying the parameters of one-step ahead predictor. Also, computer simulations are performed to evaluate the performance of proposed algorithm using a mathematical model of PWR steam generator Simulation results show good performances for load variation. And the proposed algorithm shows better responses than PI controller does.