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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
> Journal Vol & Issue
Nuclear Engineering and Technology
Journal Basic Information
Journal DOI :
Korean Nuclear Society
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Volume & Issues
Volume 29, Issue 6 - Dec 1997
Volume 29, Issue 5 - Oct 1997
Volume 29, Issue 4 - Aug 1997
Volume 29, Issue 3 - Jun 1997
Volume 29, Issue 2 - Apr 1997
Volume 29, Issue 1 - Feb 1997
Selecting the target year
Assessing the Feasibility of an Accident Management Strategy Using Dynamic Reliability Methods
Moosung Jae ; Kim, Jae-Hwan ;
Nuclear Engineering and Technology, volume 29, issue 1, 1997, Pages 1~6
This paper presents a new dynamic approach for assessing feasibility associated with the implementation of accident management strategies by the operators. This approach includes the combined use of both the concept of reliability physics and a dynamic event tree generation scheme. The reliability physics is based on the concept of a comparison between two competing variables, i.e., the requirement and the achievement parameter, while the dynamic event tree generation scheme on the continuous generation of the possible event sequences at every branch point up to the desired solution. This approach is applied to a cavity flooding strategy in a reference plant, which is to supply water into the reactor cavity using emergency fire systems in the station blackout sequence. The MAAP code and Latin Hypercube sampling technique are used to determine the uncertainty of the requirement parameter. It has been demonstrated that this combined methodology may contribute to assessing the success likelihood of the operator actions required during accidents and therefore to developing the accident management procedures.
Tracer Concentration Contours in Grain Lattice and Grain Boundary Diffusion
Kim, Yong-Soo ; Donald R. Olander ;
Nuclear Engineering and Technology, volume 29, issue 1, 1997, Pages 7~14
Grain boundary diffusion plays a significant role in fission gas release, which is one of the crucial processes dominating nuclear fuel performance. Gaseous fission produce such as Xe and Kr generated during nuclear fission have to diffuse in the grain lattice and the boundary inside fuel pellets before they reach the open spaces in a fuel rod. These processes can be studied by 'tracer diffusion' techniques, by which grain boundary diffusivity can be estimated and directly used for low burn-up fission gas release analysis. However, only a few models accounting for the both processes are available and mostly handle them numerically due to mathematical complexity. Also the numerical solution has limitations in a practical use. In this paper, an approximate analytical solution in case of stationary grain boundary in a polycrystalline solid is developed for the tracer diffusion techniques. This closed-form solution is compared to available exact and numerical solutions and it turns out that it makes computation not only greatly easier but also more accurate than previous models. It can be applied to theoretical modelings for low bum-up fission gas release phenomena and experimental analyses as well, especially for PIE (post irradiation examination).
Design and Evaluation of the Model Based Controller for a U-tube Steam Generator Level
Kim, Keung-Koo ; Lee, Doojeong ; John E. Meyer ; David D. Lanning ; John A. Bernard ;
Nuclear Engineering and Technology, volume 29, issue 1, 1997, Pages 15~24
The design and evaluation of a digital U-tube steam generator level controller of nuclear power plants, which uses model-based compensators to offset the inverse response behavior of water level, is described. Included is a review of steam generator level dynamics, a simulation model that replicates the effects of feedwater and steam flowrate as well as temperature on steam generator level, the design of both the compensators and the overall controller, and the results of simulation studies in which the performances of this model-based controller and existing analog ones were compared. The proposed digital steam generator level controller is stable and its use significantly improves the controllability of steam generator level.
Expert Opinion Elicitation Process Using a Fuzzy Probability
Yu, Donghan ;
Nuclear Engineering and Technology, volume 29, issue 1, 1997, Pages 25~34
This study presents a new approach for expert opinion elicitation process to assess an uncertainty inherent in accident management. The need to work with rare event and limited data in accident management leads analysis to use expert opinions extensively. Unlike the conventional approach using point-valued probabilities, the study proposes the concept of fuzzy probability to represent expert opinion. The use of fuzzy probability has an advantage over the conventional approach when an expert's judgment is used under limited dat3 and imprecise knowledge. The study demonstrates a method of combining and propagating fuzzy probabilities. finally, the proposed methodology is applied to the evaluation of the probability of a bottom head failure for the flooded case in the Peach Bottom BWR nuclear power plant.
An Empirical Correlation for Critical Flow Rates of Subcooled Water Through Short Pipes with Small Diameters
Park, Choon-Kyung ; Park, Jee-Won ; Chung, Moon-Ki ; Chun, Moon-Hyun ;
Nuclear Engineering and Technology, volume 29, issue 1, 1997, Pages 35~44
Critical too-Phase flow rates of subcooled water through Short Pipes (L 140039n) with small diameters (D
7.15 min) have been experimentally investigated for wide ranges of subcooling (0~199
) and pressure (0.5~2.0 MPa). To examine the effects of various parameters (i.e., the location of flashing inception, the degree of subcooling, the stagnation temperature and pressure, and the pipe size) on the critical two-phase flow rates of subcooled water through short pipes with small diameters, a total of 135 runs were made for various combinations of test parameters using four different L/D test sections. Experimental results that show effect of various parameters on subcooled critical two phase flow rates are presented in the form of graphs such as the dimensionless mass flux (
) versus the dimensionless subcooling (
) curve. An empirical correlation expressed in terms of a dimensionless subcooling is also obtained for subcooled two-phase flow rates through present test sections. Comparisons between the mass fluxes calculated by present correlation and a total of 755 selected experimental data points of 9 different investigators show that the agreement is fairly good except for very low subcooling data obtained from small L/D (less than 10) orifices.s.s.s.
An Experimental Investigation of Direct Condensation of Steam Jet in Subcooled Water
Kim, Yeon-Sik ; Chung, Moon-Ki ; Park, Jee-Won ; Chun, Moon-Hyun ;
Nuclear Engineering and Technology, volume 29, issue 1, 1997, Pages 45~57
The direct contact condensation phenomenon, which occurs when steam is injected into the subcooled water, has been experimentally investigated. Two plume shapes in the stable condensation regime are found to be conical and ellipsoidal shapes depending on the steam mass flux and the liquid subcooling. Divergent plumes, however, are found when the subcooling is relatively small. The measured expansion ratio of the maximum plume diameter to the injector inner diameter ranges from 1.0 to 2.3. By means of fitting a large amount of measured data, an empirical correlation is obtained to predict the steam plume length as a function of a dimensionless steam mass flux and a driving potential for the condensation process. The average heat transfer coefficient of direct contact condensation has been found to be in the range 1.0~3.5 ㎿/
. Present results show that the magnitude of the average condensation heat transfer coefficient depends mainly on the steam mass fin By using dynamic pressure measurements and visual observations, six regimes of direct contact condensation have been identified on a condensation regime map, which are chugging, transition region from chugging to condensation oscillation, condensation oscillation, bubbling condensation oscillation, stable condensation, and interfacial oscillation condensation. The regime boundaries are quite clearly distinguishable except the boundaries of bubbling condensation oscillation and interfacial oscillation condensation.
A Numerical Study on Mixing Characteristics of the Chemical Injection Tank
Chang, Keun-Sun ; Park, Byeong-Ho ;
Nuclear Engineering and Technology, volume 29, issue 1, 1997, Pages 58~67
A numerical study has been peformed to investigate the flow and mixing characteristics of a chemical injection tank in the chemical and volume control system (CVCS) of Yonggwang 5&6 (YGN 5&6). This study was undertaken to provide a basis for modification of the previous design (YGN 3&4) which gave a lot of difficulties in installation and operation of the chemical injection system during the start-up test because it needs a special reciprocating pump with a high actual head. For the tank of length-to-diameter ratios (L/D) of 1,2 and 3, each with and without a baffle inside, calculation results were obtained by solving the unsteady laminar two-dimensional elliptic forms of governing equations for the mass, momentum and species concentration. Finite-difference method was used to obtain discretized equations, and the SIMPLER solution algorithm, which was developed based on the staggered grid control volume, was employed for the calculation procedure. Results showed that the baffle is very effective in enhancing the mixing in the tank and that a baffle should be installed near the tank entrance in order to 110 chemicals into the reactor coolant system (RCS) within the operating time required.
Determination of the Weighting Parameters of the LQR System for Nuclear Reactor Power Control Using the Stochastic Searching Methods
Lee, Yoon-Joon ; Cho, Kyung-Ho ;
Nuclear Engineering and Technology, volume 29, issue 1, 1997, Pages 68~77
The reactor power control system is described in the fashion of the order increased LQR system. To obtain the optimal state feedback gain vectors, the weighting matrix of the performance function should be determined. Since the contentional method has some limitations, stochastic searching methods are investigated to optimize the LQR weighting matrix using the modified genetic algorithm combined with the simulated annealing, a new optimizing tool named the hybrid MGA-SA is developed to determine the weighting parameters of the LQR system. This optimizing tool provides a more systematic approach in designing the LQR system. Since it can be easily incorporated with any forms of the cost function, it also provides the great flexibility in the optimization problems.
Development of the Numerical Guide for Cost-Benefit Analysis of Occupational Radiation Exposure In the Korean Next Generation Reactor
Sohn, Ki-Yoon ; Kang, Chang-Sun ;
Nuclear Engineering and Technology, volume 29, issue 1, 1997, Pages 78~84
The specific purpose of this study is to develop the numerical guide for the cost-benefit analysis of ORE ($/person-Sv reduction) to meet the criterion of ALARA in the design stage of the KNGR. In deriving the guide, the risk factor which is defined by the risk to unit collective radiation exposure dose (deaths/person-Sv) and the monetary value of human life ($/death) are required. The risk factor has been estimated from various clinical data accumulated for a number of years and continuously modified. And the monetary value of human life is usually quantified using the human capital approach. In this study, the risk to radiation exposure perceived by a group of people is investigated through an extensive poll survey conducted among university students in order to modify the existing risk factor for radiation exposure. And in evaluating the monetary value of human life, the QOL factor is introduced in order to incorporate the degree of public welfare or quality of life. As a result of study, a value within the range of 151, 000~172, 000 dollars per person-Sv reduction is recommended as the appropriate interim numerical guide for cost-benefit analysis of ORE to meet the criterion of ALARA in the design stage of the KNGR. A poll survey was also conducted in order to see whether the public acceptance cost of nuclear power should be incorporated in developing the guide, and the result of study shooed that such a cost does not need to be considered.
Flow-Induced Vibration Test in the Preheater Region of a Steam Generator Tube Bundle
Kim, Beom-Shig ; Hwang, Jong-Keun ;
Nuclear Engineering and Technology, volume 29, issue 1, 1997, Pages 85~91
Cross-flow existing in a shell-and-tube steam generator can cause a tube to vibrate. There are four regions subjected to cross-flow in Yonggwang units 3 and 4 (YGN 3 and 4) steam generators, which are of the same design as the steam generators for Palo Verde nuclear power plant Palo Verde units 1 and 2 steam generators have experienced localized oar at the comers of the cold side recirculating fluid inlet regions. A number of design modifications were made to preclude tube failure in specific regions of YGN 3 and 4 steam generators. Therefore, flow induced vibration experiments were done to determine the vibration magnitude of tubes in the economizer tube free lane region. The objective of this experiment is to demonstrate that the tube displacement is less than 0.01 inch rms at 100% of full power flow and to quantify the remaining design margin at 120ft and 140% of full power flow.