Go to the main menu
Skip to content
Go to bottom
REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
> Journal Vol & Issue
Nuclear Engineering and Technology
Journal Basic Information
Journal DOI :
Korean Nuclear Society
Editor in Chief :
Volume & Issues
Volume 31, Issue 6 - Dec 1999
Volume 31, Issue 5 - Oct 1999
Volume 31, Issue 4 - Aug 1999
Volume 31, Issue 3 - Jun 1999
Volume 31, Issue 2 - Apr 1999
Volume 31, Issue 1 - Feb 1999
Selecting the target year
A Power Control System for the Rod Drive Coil of Control Element Drive Mechanism in Pressurized Water Reactor
Hwang, Dong-Hwan ; Seong, Se-Jin ; Park, Gwang-Seok ;
Nuclear Engineering and Technology, volume 31, issue 1, 1999, Pages 1~8
In this paper, we propose a new type of power control system for the rod drive coil of the CEDM of the PWR NPP in order to supply more reliable DC power The electrical modelling of the controlled rod drive coil was done by referring related documentations. The design of the proposed system is based on this electric81 model satisfying the existing specification. A high power DC-DC converter scheme is adopted utilizing the SMPS technique in the design of the proposed system. In order to show the effectiveness of the proposed system, an experimental system with the capability of 3.2 K Watt was set up for a rod with four cores and some computer simulations and experimentations were carried out. The result shows a very similar tracking performance with that of the existing system to the driving command. As a result of this, the proposed method can be applied to the power control system for the rod drive coil of the CEDM of the PWR NPP.
Evaluation of Radioactive Source Terms in the System-Integrated Modular Advanced Reactor
Kim, Seong-Uck ; Kang, Chang-Sun ;
Nuclear Engineering and Technology, volume 31, issue 1, 1999, Pages 9~16
A 330 MWt-sized multi-purpose integral-type reactor, SMART is under development in Korea for the use of nuclear energy other than electricity generation. In this study, various radioactive source terms are estimated for SMART. SMART is different from conventional reactor concepts in operation and design. Therefore Specific Calculation method namely recurrence model is used. This model is based on the change rate in the RC radioactivity materials and operational characteristics of SMART Calculation results show tremendously increase of the levels of RC activity because no cleanup of RC and long term operation.
Loss of Coolant Accident Analysis During Shutdown Operation of YGN Units 3/4
Bang, Young-Seok ; Kim, Kap ; Seul, Kwang-Won ; Kim, Hho-Jung ;
Nuclear Engineering and Technology, volume 31, issue 1, 1999, Pages 17~28
A thermal-hydraulic analysis is conducted on the loss-of-coolant-accident (LOCA) during shutdown operation of YGN Units 3/4. Based on the review of plant-specific characteristics of YGN Units 3/4 in design and operation, a set of analysis cases is determined, and predicted by the RELAP5/MOD3.2 code during LOCA in the hot-standby mode. The evaluated thermal-hydraulic phenomena are blowdown, break flow, inventory distribution, natural circulation, and core thermal response. The difference in thermal-hydraulic behavior of LOCA at shutolown condition from that of LOCA at full power is identified as depressurization rate, the delay in peak natural circulation timing and the loop seal clearing (LSC) timing. In addition, the effect of high pressure safety injection (HPSI) on plant response is also evaluated. The break spectrum analysis shows that the critical break size can be between 1% to 2% of cold leg area, and that the available operator action time for the Sl actuation and the margin in the peak clad temperature (PCT) could be reduced when considering uncertainties of the present RELAP5 calculation.
Experimental Study on Two-Phase Flow Parameters of Subcoolet Boiling in Inclined Annulus
Lee, Tae-Ho ; Kim, Moon-Oh ; Park, Goon-Cherl ;
Nuclear Engineering and Technology, volume 31, issue 1, 1999, Pages 29~48
Local two-phase flow parameters of subcooled flow boiling in inclined annulus were measured to investigate the effect of inclination on the internal flow structure. Two-conductivity probe technique was applied to measure local gas phasic parameters, including void fraction, vapor bubble frequency, chord length, vapor bubble velocity and interfacial area concentration. Local liquid velocity was measured by Pilot tube. Experiments were conducted for three angles of inclination; 0
. The system pressure was maintained at atmospheric pressure. The range of average void fraction was up to 10% and the average liquid superficial velocities were less than 1.3 m/sec. The results of experiments showed that the distributions of two-phase How parameters were influenced by the angle of channel inclination. Especially, the void fraction and chord length distributions were strongly affected by the increase of inclination angle, and flow pattern transition to slug flow was observed depending on the How conditions. The profiles of vapor velocity, liquid velocity and interfacial area concentration were found to be affected by the non-symmetric bubble size distribution in inclined channel. Using the measured distributions of local phasic parameters, an analysis for predicting average void fraction was performed based on the drift flux model and flowing volumetric concentration. And it was demonstrated that the average void fraction can be more appropriately presented in terms of flowing volumetric concentration.
Prediction of Safety Critical Software Operational Reliability from Test Reliability Using Testing Environment Factors
Jung, Hoan-Sung ; Seong, Poong-Hyun ;
Nuclear Engineering and Technology, volume 31, issue 1, 1999, Pages 49~57
It has been a critical issue to predict the safety critical software reliability in nuclear engineering area. For many years, many researches have focused on the quantification of software reliability and there have been many models developed to quantify software reliability. Most software reliability models estimate the reliability with the failure data collected during the test assuming that the test environments well represent the operation profile. User's interest is however on the operational reliability rather than on the test reliability. The experiences show that the operational reliability is higher than the test reliability. With the assumption that the difference in reliability results from the change of environment, from testing to operation, testing environment factors comprising the aging factor and the coverage factor are developed in this paper and used to predict the ultimate operational reliability with the failure data in testing phase. It is by incorporating test environments applied beyond the operational profile into testing environment factors. The application results show that the proposed method can estimate the operational reliability accurately.
The C Language Auto-generation of Reactor Trip Logic Caused by Steam Generator Water Level Using CASE Tools
Kim, Jang-Yeol ; Lee, Jang-Soo ;
Nuclear Engineering and Technology, volume 31, issue 1, 1999, Pages 58~67
The purpose is to produce a model of nuclear reactor trip logic caused by the steam generator water level of Wolsong 2/3/4 unit through an activity chart and a statechart and to produce C language automatically using Statechart-based Formalism and Stalemate MAGNUM toolset suggested by David Harel Formalism. It was worth attempting auto-generation of C language though we manually made Software Requirement Specification(SRS) for safety-critical software using statechart-based formalism. Most of the phases of the software life-cycle except the software requirement specification of an analysis phase were generated automatically by Computer Aided Software Engineering (CASE) tools. It was verified that automatically produced C language has high productivity, portability, and quality through the simulation.
Containment Closure Time Following the Loss of Shutdown Cooling Event of YGN Units 3&4
Seul, Kwang-Won ; Bang, Young-Seok ; Kim, Hho-Jung ;
Nuclear Engineering and Technology, volume 31, issue 1, 1999, Pages 68~79
The YGN Units 3&4 plant conditions during shutdown operation were reviewed to identify the possible event scenarios following the loss of shutdown cooling (SDC) event. For the five cases of typical reactor coolant system (RCS) configurations under the worst event sequence, such as unavailable secondary cooling and no RCS inventory makeup, the thermal hydraulic analyses were performed using the RELAP5/MOD3.2 code to investigate the plant behavior following the event. The thermal hydraulic analyses include the estimation of time to boil, time to core uncovery, and time to core heat up to determine the containment closure time to prevent the uncontrolled release of fission products to atmosphere. The result indicates that the containment closure is recommended to be achieved within 42 minutes after the loss of SDC for the steam generator (SG) inlet plenum manway open case or the large cold leg open case under the worst event sequence. The containment closure time is significantly dependent on the elevation and size of the opening and the SG secondary water level condition. It is also found that the containment closure needs to be initiated before the boiling time to ensure the survivability of the workers in the containment. These results will provide useful information to operators to cope with the loss of SDC event.
Analytic Modeling of the Xenon Oscillation Due to Control Rod Movement
Song, Jae-Seung ; Cho, Nam-Zin ; Zee, Sung-Quun ;
Nuclear Engineering and Technology, volume 31, issue 1, 1999, Pages 80~87
An analytic axial xenon oscillation model was developed for pressurized water reactor analysis. The model employs an equation system for axial difference parameters that was derived from the two-group one-dimensional diffusion equation with control rod modeling and coupled with xenon and iodine balance equations. The spatial distributions of nu, xenon, and iodine were expanded by the Fourier sine series, resulting in cancellation of the flux-xenon coupled non-linearity. An inhomogeneous differential equation system for the axial difference parameters, which gives the relationship between power, iodine and xenon axial differences in the case of control rod movement, was derived and solved analytically. The analytic solution of the axial difference parameters can directly provide with the variation of axial power difference during xenon oscillation. The accuracy of the model is verified by benchmark calculations with one-dimensional reference core calculations.
Post Test Analysis of the Phebus FPT1 Experiment
Cho, Song-Won ; Park, Jong-Hwa ; Kim, Hee-Dong ;
Nuclear Engineering and Technology, volume 31, issue 1, 1999, Pages 88~103
The purposes of this study are to understand the severe accident phenomena, to establish the simulation method for the experimental test, and to assess the current models in MELCOR for future improvement. This paper presents the results of the PHEBUS FPT1 post test analysis using MELCOR computer code, version 1.8.4. The entire PHEBUS facility has been modeled; the core, the primary circuit including the steam generator, and the containment vessel. Both the thermal hydraulic and the fission product behavior have been investigated. The code simulation results of the thermal hydraulic behavior show good agreement with the experimental data, The fission product release and transport are calculated using the CORSOR models in MELCOR code and the results will be compared with the experiment when the experimental data are available.
Prediction of the Turbulent Mixing in Bare Rod Bundles
Kim, Sin ; Chung, Bum-Jin ;
Nuclear Engineering and Technology, volume 31, issue 1, 1999, Pages 104~115
The turbulent mixing rate is a very important variable in the thermal-hydraulic design of nuclear reactors. In this study, the turbulent mixing rate the fluid flows through rod bundles is estimated with the scale analysis on the flow pulsation phenomenon. Based upon the assumption that the turbulent mixing is composed of molecular motion, isotropic turbulent motion (turbulent motion without the flow pulsation), and How pulsation, the scale relation for the mixing is derived as a function of P/D, Re, and Pr. The derived scale relation is compared with published experimental results and shows good agreements. Since the scale relation is applicable to various Prandtl number fluid flows, it is expected to be useful for the thermal-hydraulic analysis of liquid metal coolant reactors as well as of moderate Prandtl number coolant reactors.
Analysis of Functional Criteria for Buffer Material in a High-level Radioactive Waste Repository
W. J. Cho ; Lee, J. O. ; K. S. Chun ; Park, Hyun-Soo ;
Nuclear Engineering and Technology, volume 31, issue 1, 1999, Pages 116~132
This study is intended to analyze the requirements of a buffer material that is one of the major components of the engineered barriers in a high-level radioactive waste repository. The characteristics of potential materials for the buffer in the repository were analyzed and a candidate material was selected. And, based on the current knowledge and the information from various sources, the requirements of a buffer material were evaluated. Finally its quantitative functional criteria on the generic viewpoint has been recommended to be supplied as a guideline for the development of the reference disposal concept and the related buffer material in Korea. The criteria are composed of seven major items, such as hydraulic conductivity, retardation capacity, swelling potential and swelling pressure, thermal conductivity, longevity, organic matter content, and mechanical properties.