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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
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Journal DOI :
Korean Nuclear Society
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Volume & Issues
Volume 25, Issue 4 - Dec 1993
Volume 25, Issue 3 - Sep 1993
Volume 25, Issue 2 - Jun 1993
Volume 25, Issue 1 - Mar 1993
Selecting the target year
Finite Element Analysis of Pipe Whip Restraint Behavior Under Jet Thrust Forces
Sugoong Koh ; Lee, Young-Shin ;
Nuclear Engineering and Technology, volume 25, issue 3, 1993, Pages 353~360
Many types of pipe whip restraints are installed to protect the structural components from the anticipated pipe whip phenomena of high energy lines in nuclear power plants. It is necessary to investigate these phenomena accurately in order to design the pipe whip restraints properly and/or to evaluate the acceptability of the pipe whip restraint design. Various research programs have been conducted in many countries to develop analytical methods and to verify the validity of the methods. In this study, various types of finite elements in ANSYS, the general purpose finite element computer program, was used to simulate the postulated pipe whips to obtain impact loads and the calculated results were compared with the specific experimental results from the sample pipe whip test for the U-shaped pipe whip restraints. Some calculational models, having the gap element or the spring element between the pipe whip restraint and the pipe line, give reasonably good transient responses of the restraint forces compared with the experimental results, and could be useful in evaluating the acceptability of the pipe whip restraint design.
A Study on the Effect of Gamma Background in Low Power Startup Physics Tests
Bae, Chang-Joon ; Lee, Ki-Bog ;
Nuclear Engineering and Technology, volume 25, issue 3, 1993, Pages 361~370
Low power physics tests should be peformed for the domestic pressurized light water reactors (PWRs) after refueling. The tests are peformed to ensure that operating characteristics of the core are consistent with predictions and that the core can be operated as designed. But in some low power physics tests, slow but steady reactivity increasing phenomena were noticed after step reactivity insertion by the control rod movement. These reactivity increasing phenomena are due to the low flux level and the gamma background because an uncompensated ion chamber (UIC) is used as the ex-core neutron detector. The gamma background may affect the results or the lour power physics tests. The aims or this paper are to analyze the grounds of such phenomena, to simulate a reference bank worth measurement test and to present a resolution quantitatively. In this study, the gamma background level was estimated by numerically solving the point kinetics equations accounting the gamma background effect. The reactivity computer check test was simulated to verify the model. Also, an appropriate neutron flux level was determined by simulating the reference bank worth measurement test. The determined neutron flux level is approximately 0.3 of the nuclear heating flux. This level is about 3 times as high as the current test upper limit specified in the test procedure. Then, the findings from this work were successfully applied to Kori unit 4 cycle 7 and Yonggwang unit 1 cycle 7 physics tests.
A 9-Rule Fuzzy Logic Controller of the Nuclear Steam Generator
Lee, Jae-Young ; No, Hee-Cheon ;
Nuclear Engineering and Technology, volume 25, issue 3, 1993, Pages 371~380
A model free controller utilizing a set of linguistic fuzzy logic of the human operator's experience is developed to control the steam generator water level in a pressurized water reactor. Only 9 rules for control action are generated from the inputs of water level error and mass flow error implicitly representing the time variation of the collapsed water level. The bell type membership functions of the premise side and the result side are tuned by the sensitivity study. This compact fuzzy logic controller shows a robust control during transient and no offset error and oscillation during steady state operation. For a multi-ramp power increase from start-up to full power, the proposed controller shows good performance for the entire range.
Air-Water Flooding in Multirod Channels : Effects of Spacer Grids and Blockages
Cha, Jong-Hee ; Jun, Hyung-Gil ;
Nuclear Engineering and Technology, volume 25, issue 3, 1993, Pages 381~393
This paper presents the experimental results on flooding of countercurrent flow in vertical multirod channels, which consists of falling water film and upward air flow. In particular, the effects of spacer grids, with and without mixing vane, and of blockage in the multirod bundle on the behaviour of flooding were investigated. The 5
5 zircaloy tube bundle was used for the test section. The comparison of previous analytical models and empirical correlations with present data on flooding showed that the existing models and correlations predict much higher flooding curves. The spacer grid causes the lower flooding air flow rate to compare with the bare rod bundle. However, the mixing spacer grids need a higher flooding air flow rate for a constant liquid flow rate than the spacer grids without mixing vanes. The bundle containing blockages has the highest flooding air flow rate among the bundles with spacer grids and blockages. Empirical flooding correlations for the three types of test section have been made.
Model Developments for Quantitative Estimates of the Benefits of the Signals on Nuclear Power Plant Availability and Economics
Seong, Poong-Hyun ;
Nuclear Engineering and Technology, volume 25, issue 3, 1993, Pages 394~402
A novel framework for quantitative estimates of the benefits of signals on nuclear power plant availability and economics has been developed in this work. The models developed in this work quantify how the perfect signals affect the human operator's success in restoring the power plant to the desired state when it enters undesirable transients. Also, the models quantify the economic benefits of these perfect signals. The models have been applied to the condensate feedwater system of the nuclear power plant for demonstration.
Application of Smart Transmitter Technology in Nuclear Engineering Measurements
Kang, Hyun-Gook ; Seong, Poong-Hyun ;
Nuclear Engineering and Technology, volume 25, issue 3, 1993, Pages 403~412
By making use of the microprocessor technology, instrumentation system becomes intelligent. In this study a programmable smart transmitter is designed and applied to the nuclear engineering measurements. In order to apply the smart transmitter technology to nuclear engineering measurements, the digital time delay compensation function and water level change detection function are developed and applied in this work. The time compensation function compensates effectively the time delay of the measured signal, but it is found that the characteristics of the compensation function should be considered through its application. It is also found that the water level change detection function reduces the detection time to about 7 seconds by the signal processing which has the time constant of over 250 seconds and which has the heavy noise.
An Experimental Study of Direct Containment Heating Phenomena
Chanyoung Chung ; Gyoodong Jeun ; Bang, Kwang-Hyun ; Kim, Moohwan ;
Nuclear Engineering and Technology, volume 25, issue 3, 1993, Pages 413~423
This paper reports an experimental study of direct containment heating (DCH) which would occur if the primary system pressure is still high at the time of vessel breach during a light water reactor core melt accident. The experiments were conducted in 1/30-scale cavity models of Kori unit 1 and 2 and Young Kwang unit 3 and 4 nuclear power plants. One 1/20-scale model of the Kori plant was also used to investigate the scaling effect. The primary variables in the experiments were initial vessel pressure, vessel breach size and cavity geometry. It is observed that higher initial pressure and larger breach size enhance the melt dispersal fraction. Also, the cavity geometry appears to affect the dispersal rate greatly. A simple correlation of melt dispersal fraction is proposed in terms of nondimensional effective period. This correlation shows good agreement with the present experimental data, the KAIST data and the BNL data.
Development of Automatic Nuclear Fuel Rod Character Recognition System Based on Image Processing Technique
Woong Ki Kim ; Yong Bum Lee ; Jong Min Lee ; Sung IL Chien ;
Nuclear Engineering and Technology, volume 25, issue 3, 1993, Pages 424~429
Numeric characters are printed at the end part of nuclear fuel rod containing nuclear pellets. Fuel rods are discriminated and managed systematically by these characters in the process of producing fuel assembly. The characters are also used to examine manufacturing process of fuel rods in the survey of burnup efficiency as well as in inspection of irradiated fuel rod. Therefore automatic character recognition is one of the most important technologies in automatic manufacture of fuel assembly. In this study, character recognition system is developed. In the developed system, mesh feature extracted from each character written in the fuel rod has been compared with reference feature value stored in database, and the character is thus identified. In the result of experiment, 95.83 percent recognition rate is achievable.
A Study on the Optimal Replacement Periods of Digital Control Computer's Components of Wolsung Nuclear Power Plant Unit 1
Mok, Jin-Il ; Seong, Poong-Hyun ;
Nuclear Engineering and Technology, volume 25, issue 3, 1993, Pages 430~436
Due to the failure of the instrument and control devices of nuclear power plants caused by aging, nuclear power plants occasionally trip. Even a trip of a single nuclear power plant (NPP) causes an extravagant economical loss and deteriorates public acceptance of nuclear power plants. Therefore, the replacement of the instrument and control devices with proper consideration of the aging effect is necessary in order to prevent the inadvertent trip. In this paper we investigated the optimal replacement periods of the control computer's components of Wolsung nuclear power plant Unit 1. We first derived mathematical models for optimal replacement periods to the digital control computer's components of Wolsung NPP Unit 1 and calculated the optimal replacement periods analytically. We compared the periods with the replacement periods currently used at Wolsung NPP Unit 1. The periods used at Wolsung is not based on mathematical analysis, but on empirical knowledge. As a consequence, the optimal replacement periods analytically obtained and those used in the field show a little difference.
Improvement on the KFOOD Code for More Realistic Assessment of the Annual Food Chain Radiation Dose Due to Operating Nuclear Facilities
Park, Yong-Ho ; Lee, Chang-Woo ; Kim, Jin-Kyu ; Lee, Myung-Ho ; Lee, Jeong-Ho ;
Nuclear Engineering and Technology, volume 25, issue 3, 1993, Pages 437~446
More realistic calculation models for evaluating man's annual intakes of radionuclides released from operating nuclear facilities were established. For the application of these models, the harvest years of food and feed crops consumed in the year of dose assessment and every year's average concentrations of a radionuclide in air and in water for the whole period of real operation had to be taken into account. KFOOD, an existing equilibrium food chain computer code for the Korean dose assessment, was modified according to the models. Sample runs of the modified code on the assumption of a constant release during 10 years' operation were made with three kinds of the input data files enabling the dose assessment in the improved method, the KFOOD method and another existing method, respectively, and the results were compared. Annual committed effective doses to Korean adult by intakes of Mn-54, Co-60, Sr-90, I-131 and Cs-137 calculated in the improved method were about 11, 2, 5, 60 and 3%, respectively, lower than the corresponding KFOOD dose. To the intakes of the radionuclides except Sr-90 evaluated in the improved method, foliar uptake contributed much more than root uptake did but, in the case of Sr-90, the result was opposite.
Performance Assessment of Engineered Barrier for Retardation of Radionuclide Release in a Low- and Intermediate-Level Radioactive Waste Repository
W.J. Cho ; Lee, J.O. ; P.S. Hahn ; Park, H.H. ;
Nuclear Engineering and Technology, volume 25, issue 3, 1993, Pages 447~456
A simplified model to assess the performance of engineered barrier for the retardation of radionuclide release in a low- and intermediate-level radioactive waste repository was developed. The model is based on the repository design concept being suggested in Korea, and considers two types of release scenario ; a design-bas release for the design of engineered barrier and a realistic release for the performance assessment. For the numerical illustration, the sample calculations were performed for five radionuclides with different chemical characteristics, and the results were analyzed.
An Analysis of Fast Critical Experiments Using JEF-1-Based 50-Group Constant Set
Kim, Jung-Do ; Gil, Choong-Sup ; Kim, Young-Cheol ;
Nuclear Engineering and Technology, volume 25, issue 3, 1993, Pages 457~469
JEF-1-based 50-group cross section set for fast reactor calculations was generated using NJOY system. The set was then examined by analyzing measured integral quantities such as criticality and central reaction rate ratios for 27 fast critical assemblies. The calculated results using the new set were also compared with those of ENDF/B-IV or-V-based fast set. In general, the JEF-1-based set shows an improvement in predicting measured integral quantities in comparison with the previous set. With a few exceptions, JEF-1 results are comparable to those of ENDF/B-V.
영광 2호기 4주기 운전자료 비교 분석
Nuclear Engineering and Technology, volume 25, issue 3, 1993, Pages 470~479
고준위 방사성핵종 소멸처리 기술의 검토 -핵특성 관점에서-
Nuclear Engineering and Technology, volume 25, issue 3, 1993, Pages 480~496
IAEA제정 [신형원자로의 안전관련용어] 소개
Nuclear Engineering and Technology, volume 25, issue 3, 1993, Pages 497~506