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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
Journal Basic Information
Journal DOI :
Korean Nuclear Society
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Volume & Issues
Volume 30, Issue 6 - Dec 1998
Volume 30, Issue 5 - Oct 1998
Volume 30, Issue 4 - Aug 1998
Volume 30, Issue 3 - Jun 1998
Volume 30, Issue 2 - Apr 1998
Volume 30, Issue 1 - Feb 1998
Selecting the target year
Seismic Response Analyses of Seismically Isolated Structures Using the Laminated Rubber Bearings
Koo, Gyeong-Hoi ; Lee, Jae-Han ; Bong Yoo ;
Nuclear Engineering and Technology, volume 30, issue 5, 1998, Pages 387~395
In general, the laminated rubber bearing (LRB), a composite structure laminated with the elastic rubber and steel plates, has a complex hysteretic nonlinear characteristics in relationships between the restoring force and shear deflection. The representative nonlinear characteristics of LRB include the change of hysteresis loop with cyclic shear deflections and the hardening effects at large shear deflection regions. Changes of the hysteresis loop of LRB with cyclic shear deflections affect the horizontal stiffness and the damping characteristics. The hardening behavior of LRB in large shear deflection region results in an increased horizontal stiffness and therefore, has a great impacton the seismic responses. In this paper, the seismic response analysis is carried out using the modified hysteretic bi-linear model of LRB, which takes into account the hysteresis loop change and the hardening behavior with cyclic shear deflection. The results on seismic responses are compared with those obtained using the widely used hysteretic hi-linear model. The new model is found to reveal the greater amount of peak acceleration response.
Pressure Effects on Zircaloy-4 Steamside Corrosion and Hydrogen Pick-up
Ok, Young-kil ; Kim, Yong-soo ;
Nuclear Engineering and Technology, volume 30, issue 5, 1998, Pages 396~402
Experiments on the steamside corrosion and hydrogen pick-up of Zircaloy-4 under high pressure up to 10.3MPa are carried out to estimate the pressure effects on the kinetics. Temperature and reaction time are determined to be 37
and 72hours for the pre-transition test and
and 210minutes for the post-transition test, respectively. Results show that under 10.3MPa pressure the oxidation reaction is 50% and 100% enhanced in the pre-and the post-transition regime, respectively. Total amount of hydrogen uptake in the reaction is proportionally increased as corrosion weight gain is elevated. However, pick-up fraction is not affected by the high pressure. The fraction is almost twice greater than that in the waterside corrosion. Edges in the specimens play a certain role in the enhancement, especially in the post-transition regime. To identify physical property changes of oxide film such as micro-cracks or micro-pores, careful and thorough examination must be needed with some special techniques.
Assessment of Leak Detection Capability of CANDU 6 Annulus Gas System Using Moisture Injection Tests
Nho, Ki-Man ; Kim, Wang-Bae ; Sim, Woo-Gun ;
Nuclear Engineering and Technology, volume 30, issue 5, 1998, Pages 403~415
The CANDU 6 reactor assembly consists of an array of 380 pressure tubes, which are installed horizontally in a large cylindrical vessel, the Calandria, containing the low pressure heavy water moderator. The pressure tube is located inside the calandria tube and the annulus between these tubes, which forms a closed loop with
gas recirculating, is called the Annulus Gas System(AGS). It is designed to give an alarm to the operator even for a small pressure tube leak by a very sensitive dew point meter so that he can take a preventive action for the pressure tube rupture incident. To judge whether the operator action time is enough or not in the design of Wolsong 2,3 & 4, the Leak Before Break(LBB) assessment is required for the analysis of the pressure tube failure accident. In order to provide the required data for the LBB assessment of Wolsong Units 2, 3, 4, a series of leak detection capability tests was performed by injecting controlled rates of heavy water vapour. The data of increased dew point and rates of rise were measured to determine the alarm set point for the dew point rate of rise of Wolsong Unit 2. It was found that the response of the dew point depends on the moisture injection rate,
gas flow rate and the leak location. The test showed that CANDU 6 AGS can detect the very small leaks less than few g/hr and dew point rate of rise alarm can be the most reliable alarm signal to warn the operator. Considering the present results, the first response time of dew point to the AGS
flow rate is approximated.
Free Vibration Analysis of Perforated Plates Using Equivalent Elastic Properties
Park, Suhn ; Jeong, Kyeong-Hoon ; Kim, Tae-Wan ; Kim, Kang-Soo ; Park, Keun-Bae ;
Nuclear Engineering and Technology, volume 30, issue 5, 1998, Pages 416~423
Many studies for the perforated plates have been done, especially on the subject of static behavior and stress distribution in the plate. Equivalent elastic properties are one of the successive concepts for this problem. However little effort was taken to get their dynamic characteristics. In this paper finite element modal analysis was performed for the perforated plates having square and triangular hole patterns. An attempt to use existing equivalent elastic properties into the modal analysis of the plate was carried out. To verify feasibility of the finite element models, modal test was also performed on one typical perforated plate. System parameters such as natural frequencies and mode shapes were extracted and compared with the analysis results.
The Simulation of Semicale Natural Circulation Test 5-NC-3,S-NC-4 Using RELAP5/Mod3.1
Kim, S. N. ; W. H. Jang ;
Nuclear Engineering and Technology, volume 30, issue 5, 1998, Pages 424~434
RELAP5/Mod3.1 code was assessed with the semiscale experiment S-NC-3, and S-NC-4, which simulated the two-phase natural circulation and reflux condensation for the SBLOCA of PWR, respectively . Test S-NC-3 and S-NC-4 calculation results showed that RELAP5/Mod3.1 quite well describes the influence of steam generator secondary side heat transfer degradation on both two-phase natural circulation and reflux condensation. A comparison between the calculated and measured two-phase mass flow rate in test S-NC-3 shows good agreement for primary mass inventory more than 92%. And RELAP5/Mod3.1 have a good mass flow rate prediction capability for the transient such as S-NC-4 except some flow oscillations. The reflux flow rate for S-NC-4 test is under predicted, and the overall results verify that the correct prediction of the reduced liquid level appears to be required for the correct calculation of the overall phenomena.
Development of Ceramic Humidity Sensor for the Korean Next Generation Reactor
Lee, Na-Young ; Hwang, Il-Soon ; Song, Chang-Rock ; Yoo, Han-Ill ; Park, Sang-Duk ; Yang, Jun-Seong ;
Nuclear Engineering and Technology, volume 30, issue 5, 1998, Pages 435~443
Leak-before-break(LBB) approach has been shown to be both cost effective and risk reductive when applied to high energy Piping in nuclear Power Plants. For the Korean Next Generation Reactor (KNGR) development, LBB application is considered for the Main Steam Line(MSL) piping inside containment. Unlike the primary system leakages, the MSL leak detection systems must be based on principles other than radioactivity measurements. Among humidity, heat and acoustic noise currently being considered as indicators of leakage, we explored humidity as an effective one and developed ceramic-based humidity sensor which can be qualified for LBB applications. The ceramic material, sintered and annealed MgCr
, is shown to increase its electrical conductivity drastically upon water vapor adsorption over the entire temperature range of interest. With this ceramic sensor specimen, we suggested installation-inside-the-piping method by which we can detect leakage more rapidly and sensitively. In this paper, we describe the progress in the development and characterization of ceramic humidity sensor for the LBB application to the MSL of KNGR.
Use of Rank Sum Method in Identifying High Occupational Dose Jobs for ALARA Implementation
Cho, Yeong-Ho ; Kang, Chang-Sun ;
Nuclear Engineering and Technology, volume 30, issue 5, 1998, Pages 444~451
The cost-effective reduction of occupational radiation exposure (ORE) dose at a nuclear power plant could not be achieved without going through an extensive analysis of accumulated ORE dose data of existing plants. It is necessary to identify what are high ORE jobs for ALARA implementation. In this study, the Rank Sum Method (RSM) is used in identifying high ORE jobs. As a case study, the database of ORE-related maintenance and repair jobs for Kori Units 3 and 4 is used for assessment, and top twenty high ORE jobs are identified. The results are also verified and validated using the Friedman test, and RSM is found to be a very efficient way of analyzing the data.
Pin Power Distribution Determined by Analyzing the Rotational Gamma Scanning Data of HANARO Fuel Bundle
Lee, Jae-Yun ; Park, Hee-Dong ;
Nuclear Engineering and Technology, volume 30, issue 5, 1998, Pages 452~461
The pin power distribution is determined by analyzing the rotational gamma scanning data for 36 element fuel bundle of HANARO. A fission monitor of Nb
is chosen by considering the criteria of the half-life, fission yield, emitting
-ray energy and probability. The
-ray spectra were measured in Korea Atomic Energy Research Institute(KAERI) by using a HPGe detector and by rotating the fuel bundle at steps of 10
. The counting rates of Nb
-rays are determined by analyzing the full absorption peak in the spectra. A 36
36 response matrix is obtained from calculating the contribution of each rod at every scanning angle by assuming 2-dimensional and parallel beam approximations for the measuring geometry. In terms of the measured counting rates and the calculated response matrix, an inverse problem is set up for the unknown distribution of activity concentrations of pins. To select a suitable solving method, the performances of three direct methods and the iterative least-square method are tested by solving simulation examples. The final solution is obtained by using the iterative least-square method that shows a good stability. The influences of detection error, step size of rotation and the collimator width are discussed on the accuracy of the numerical solution. Hence an improvement in the accuracy of the solution is proposed by reducing the collimator width of the scanning arrangement.
Structural Evaluation on the Impact of a Radioisotope Package
Chung, Sung-Hwan ; Lee, Heung-Young ; Ku, Jeong-Hoe ; Seo, Ki-Seog ; Han, Hyun-Soo ;
Nuclear Engineering and Technology, volume 30, issue 5, 1998, Pages 462~469
A package to transport high-level radioactive materials is required to withstand normal transport and hypothetical accident conditions pursuant to the IAEA and domestic regulations. The package should maintain the structural safety not to release radioactive material in any condition. The structural safety of the package has been evaluated by tests using proto-type or scaled-down models, however, the method by analysis is gradually utilized due to recent advancement of computers and computer codes. In this paper, to evaluate the structural safety of a radioisotope package of the KAERI, the three dimensional impact analyses under 9m free drop and 1m puncture were performed with an explicit finite-element code, the LS-DYNA3D code. The maximum stress intensity on each part was calculated and the structural safety of the package was evaluated in accordance with the regulations.
Deterministic Fracture Mechanics Analysis of Pressurized Thermal Shock
M. J. Jhung ; Park, Y. W. ;
Nuclear Engineering and Technology, volume 30, issue 5, 1998, Pages 470~484
An analysis program for the evaluation of pressure vessel integrity under pressurized thermal shock (PTS) is developed. For given material properties and transient history such as temperature and pressure, the stress distribution is calculated and then stress intensity factors are obtained for a wide range of crack sizes. The stress intensity factors are compared with the fracture toughness to check if cracking is expected to occur during the transient. Using this program a round robin problem of PTS during a small break loss of coolant transient has been analyzed as a part of the international comparative assessment study. The allowable maximum reference nil-ductility transition temperatures are determined for various crack sizes.