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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 18, Issue 4 - Dec 1986
Volume 18, Issue 3 - Sep 1986
Volume 18, Issue 2 - Jun 1986
Volume 18, Issue 1 - Mar 1986
Selecting the target year
Analysis of Loss of Offsite Power Transient Using RELAP5/MOD1/NSC; II: KNU1 Design-Base Simulation
Kim, Hyo-Jung ; Chung, Bub-Dong ; Lee, Young-Jin ; Kim, Jin-Soo ;
Nuclear Engineering and Technology, volume 18, issue 3, 1986, Pages 175~182
The KNUI (Korea Nuclear Unit 1) loss of offsite power transient as a design-base accident has been simulated using the RELAP5/MOD1/NSC computer code. The analysis is carried out using the best-estimate methodology, but the sequence and its assumptions are based on the evaluation methodology th at emphasizes conservatism. Important thermal-hydraulic parameters such as average temperature, steam generator level and pressurizer water volume are compared with the results in the KNU1 Final Safety Analysis Report (FSAR). The present analysis gives much lower RCS average temperature and pressurizer water volume, and much higher S/G water volume at the turnaround point, which may be considered to be additional improved safety margins. This is expected since the present analysis deals with the best-estimate thermal-hydraulic models as well as the initial conditions on a best-estimate basis. These additional safety margins may contribute to further validate the safety of the KNU1 in this type of accidents(Decrease in Heat Removal by the Secondary System).
Development of Pressure Drop Model for the Compartment in Reactor Containment
Park, Cheol ; Song, In-ho ; Lee, Un-Chul ;
Nuclear Engineering and Technology, volume 18, issue 3, 1986, Pages 183~193
Full scale HDR containment experiment series pointed out that the previous containment analysis models have a number of shortcomings. One of them is on the calculational model of short term (0~2sec) pressure difference. The pressure differences between subcompartments are dependent on the flow rate, fluid density, head loss coefficient, and flow area ratio. It, however, is not known that any of them is largely attributed to the disagreement of pressure difference between the measured and the calculated values. In this study, the head loss coefficients are expressed with another form to improve the analytic model. The pressure and the pressure difference are evaluated by using COMPARE code with new correlation, and the results show better agreements with experimental values for V.42 test, but overestimate the measured values for V, 43 and underestimate for V.44.
The Behavior of Microamounts of Americium in Aqueous Solution
Jae, Won-Mok ;
Nuclear Engineering and Technology, volume 18, issue 3, 1986, Pages 194~199
The behavior of micro amount of Am in aqueous solution were investigated with centrifugation method as a function of pH. In the studies described here, equilibration times were extended to 2-3 weeks to know the aging effect in radiocolloid formation. Also, the effect of the addition of foreign materials, e. g. silica gel and Fe
were examined as well as the effect of presence of concentrated electrolyte. In the results, Am appeared to be rapidly adsorbed on to impurity particles for pH < 6 and probably on the container walls by an ionic sorption process. The addition of foreign material increased the fraction of Am while the addition of concentrated electrolyte hindered the process. For pH > 7 Am behaved quite differently than for pH < 6. There appeared to be rapid sorption of some Am from solution probably on the container walls followed by partial desorption that occurred over a period of 1-2 days.s.
LOCA Analysis and Development of a Simple Computer Code for Refill-Phase Analysis
Ree, Hee-Do ; Park, Goon-Cherl ; Kim, Hyo-Jung ; Kim, Jin-Soo ;
Nuclear Engineering and Technology, volume 18, issue 3, 1986, Pages 200~208
The loss of coolant accident based on a double-ended cold leg break is analyzed with the discharge coefficient (Ca) of 0.4. This analysis covers the whole transient period from the start of depressurization to the complete refilling of the core by using RELAP4/MOD6-EM and RELAP4/ MOD6-HOT CHANNEL for the system thermal-hydraulics and the fuel performance during the blowdown phase respectively, and RELAP4/MOD6-FLOOD and TOODEE2 during the reflood phase. A simple analytical method has been developed to account for the lower plenum filling by approximating steam-water countercurrent flows and superheated wall effects at the downcomer during the refill period. Based on the informations. at the time of EOB (end-of-bypass), the refill duration time and the initial reflooding temperature were estimated and compared with the results from the RELAP4/MOD6, resulting in a good agreement. In addition, some parametric studies on the EOB were performed. The form loss coefficient between upper head and upper downcomer was found to be sensitive to the occurrence of the spurious EOB. Appropriate form loss coefficients should be taken into account to avoid the flow oscillations at the downcomer. The analyses with the six and three volume core nodalizations, respectively, show much similar trends in the system thermal-hydraulic performance, but the former case is recommended to obtain good results.
Analysis of Nuclear Power Plant Load Follow Operation by Temperature Reduction Method
Park, Sang-Yoon ; Park, Goon-Cherl ; Lee, Un-Cherl ; Kang, Chang-Sun ; Kim, Chang-Hyo ; Chung, Chang-Hyun ;
Nuclear Engineering and Technology, volume 18, issue 3, 1986, Pages 209~217
The inlet coolant temperature reduction technique has been used to extend the load follow operation further in the end-of-cycle-life(EOL). In order to simulate the technique and calculate the nuclear characteristics of a PWR core according to the load follow operation, the three dimensional computing system has been established. The analysis was performed in both MINB and SPINR modes of typical 12-3-6-3 locad follow operation for the EOL of KNU-1 plant. Moreover, the capability of return-to-power has been also tested for those two modes with the system analysis by the RETRAN-02 code. The results show that it has no difficulty to extend the load follow operation further in the EOL by applying the inlet coolant temprature reduction, and also the spinning reserve capacity(SRC) increases by 13% in MINB mode and 14% in SPINR mode Bore that used by control rods only, for 14
F drop in the inlet temperature.
Deformation Characteristics of Zircaloy-4 Fuel Cladding due to Oxidation in Environment of High Temperature and Steam
Jung, Sung-Hoon ; Suh, Kyung-Soo ; Kim, In-Sup ;
Nuclear Engineering and Technology, volume 18, issue 3, 1986, Pages 218~227
Studies were conducted to determine the extent of oxidation and same of the mechanical property changes of Zircaloy-4 fuel cladding after it was exposed to hot steam environment. The purpose of these tests was to provide some informations on the embrittlement behavior of CANDU type fuel cladding, which could be experienced under the loss-of-coolant accident conditions. The Zircaloy fuel cladding tubes were exposed in a steam environment at the temperature of 90
. The growth of the ZrO
layer combined with an oxygen rich
-phase layer into the Zircaloy tube material was found as a function of time t and temperature of steam exposure, E=1.1√Dt+0.002 where D is a temperature dependent diffusion coefficient. The tensile strength of the specimens exposed for a short period increased but decreased continuously with further exposure. The circumferential elongation was drastically changed with the exposure time while the hoop strength did't decrease greatly. The X-ray measurement of preferred orientation of the Zircaloy tube material indicated that grains in the as received tube were oriented such that the poles of the basal (0001) planes were predominantly radial, while the poles of the basal plane in the tube materials heattreated at 1,00
were oriented tangentially. It appears that this reoriented texture may contribute to lessening the decrease of the hoop strength of the heat treated Zircaloy tube material.
A Comparative Study on the Fault Diagnosis Using Fuzzy Set Concept
Hwang, Won-Guk ;
Nuclear Engineering and Technology, volume 18, issue 3, 1986, Pages 228~237
This paper provides a comparative study on methodologies for solutions of the inverse problems of certain basic fuzzy relational equations, with which fuzzy set is defined as mapping from sets into complete Brouwerian lattice. Three different algorithms developed so far are discussed and applied to fault diagnosis problem for the main coolant pump of nuclear power plants.
A Review on the Heavy Water Management Program and Operating Experience at WOLSUNG N.P.P.
Nuclear Engineering and Technology, volume 18, issue 3, 1986, Pages 238~246
이산화우라늄 핵연료의 원자로내 조사거동
Nuclear Engineering and Technology, volume 18, issue 3, 1986, Pages 247~260