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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
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Korean Nuclear Society
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Volume & Issues
Volume 35, Issue 6 - Dec 2003
Volume 35, Issue 5 - Oct 2003
Volume 35, Issue 4 - Aug 2003
Volume 35, Issue 3 - Jun 2003
Volume 35, Issue 2 - Apr 2003
Volume 35, Issue 1 - Feb 2003
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Experiments and MAAP4 Assessment for Core Mixture Level Depletion After Safety Injection Failure During Long-Term Cooling of a Cold Leg LB-LOCA
Kim, Y. S. ; B. U. Bae ; Park, G. C. ; K. Y. Sub ; Lee, U. C . ;
Nuclear Engineering and Technology, volume 35, issue 2, 2003, Pages 91~107
Since DBA(Design Basis Accidents) has been studied rather separately from SA(Severe Accidents) in the conventional nuclear reactor safety analysis, the thermal hydraulics during transition between DBA and SA has not been identified so much as each accident itself. Thus, in this study, the thermal hydraulic behavior from DBA to the commencement of SA has been experimentally and analytically investigated for the long-term cooling phase of LB-LOCA(Large-Break Loss-of-Coolant Accident). Experiments were conducted for both cases of the loop seal open and closed in an integral test loop, named as SNUF (Seoul National University Facility), which was scaled down to l/6.4 in length and 1/178 in area of the APR1400 (Advanced Power Reactor 1400MWe). The core mixture level was a main measured value since it took major role in the fuel heat-up rate, the location of fuel melting initiation and the channel blockage by melting material during SA. Experimental results were compared to MAAP4.03 to assess its model of calculating the core mixture level. MAAP4.03 overestimates the core two- phase mixture level because sweep-out and spill-over and the measures to simulate the status of loop seal are not included, which is against the conservatism. Thus, it is recommended that MAAP4.03 should be improved to simulate the thermal hydraulic phenomena, such as sweep-out, spill-over and the status of loop seal.
Uncertainty in Scenarios and Its Impact on Post Closure Long Term Safety Assessment in a Potential HLW Repository
Y.S. Hwang ; Kim, S-K ; Kang, C-H ;
Nuclear Engineering and Technology, volume 35, issue 2, 2003, Pages 108~120
In assessing the long term post closure radiological safety assessment of a potential HLW repository in Korea, three categories of uncertainties exist. The first one is the scenario uncertainty where series of different natural events are translated into written statements. The second one is the modeling uncertatinty where different mathematical models are applied for an identical scenario. The last one is the data uncertainty which can be expressed in terms of probabilistic density functions. In this analysis, three different scenarios are seleceted; a small well scenario, a radiolysis scenario, and a naturally discharged scenario. The MASCOT-K and the AMBER, probabilistic safety assessment codes based on connection of sub-modules and a compartment theory respectively, are applied to assess annual individual doses for a generic biosphere. Results illustrate that for a given scenario, predictions from two different codes fairly match well each other But the discrepancies for the different scenarios are significant. However, total doses are still well below the guideline of 2 mRem/yr. Detailed analyses with model and data uncertainties are underway to further assure the safety of a Korean reference dispsoal concept.
Evaluation of New Design Concepts for Steam Generators in Sodium Cooled Liquid Metal Reactors
Kim, Seong-O. ; Sim Yoonsub ; Kim, Eui-kwang. ; Myung-Hwan.Wi ; Han, Dohee. ;
Nuclear Engineering and Technology, volume 35, issue 2, 2003, Pages 121~132
To reduce the construction cost and enhance the safety of sodium cooled liquid metal reactors, various kinds of new design concepts were evaluated using the KALIMER operation condition. The required equipment sizes were set for plant electricity output to be similar to that of KALIMER. The evaluations were made focusing on the plant performance and implementation practicality. Each design concept was evaluated for the concept itself and design impacts to interfacing systems. Through the evaluation of the concepts, it was found that the most favorable design concept is the integrated steam generator with forced convection using lead bismuth as the intermediate heat transfer fluid between the primary sodium tube and feed water/steam tube in the steam generator.
A Study on the Stem Coefficient of Friction of Motor- operated Gate/Globe halves
Jeoung, Rae-Hyuck ; Park, Sung-Keun ; Lee, Do-Hwan ; Kim, Yang-Seok ;
Nuclear Engineering and Technology, volume 35, issue 2, 2003, Pages 133~143
Stem-stem nut coefficient of friction(COF) in motor-operated gate/globe valves is one of the important factors which determine the performance of the valve/actuators. The COF is affected greatly by the type and condition of the stem-stem nut lubricants, environmental parameters, surface condition of the stem/stem-nuts, and the number of strokes after the lubrication. In this paper, the measured data of the COFs at stem threads of some safety-related motor-operated gate/globe valves in domestic nuclear power plants are presented. In addition, the performance of the lubricants is evaluated by comparing the COFs among those valves. The results show that the measured COF at torque switch trip are higher than the unwedging COF and conservatively applicable to the unwedging COF. It is also shown that the lubricating performance based on the measured COFs varies with the lubricants.
Probabilistic Estimation of LMR Fuel Cladding Performance Under Transient Conditions
Kwon, Hyoung-Mun ; Lee, Dong-Uk ; Lee, Byung-Oon ; Kim, Young ll ; Kim, Yong-Soo ;
Nuclear Engineering and Technology, volume 35, issue 2, 2003, Pages 144~153
The object of this paper is the probabilistic failure analysis on the cladding performance of WPF(Whole Pin Furnace) test fuel pins under transient conditions, and analysis of the KALIMER fuel pin using the preceding analysis. The cumulative damage estimation and Weibull probability estimation of WPF test are performed. The probabilistic method was adapted for these analyses to determine the effective thickness thinning due to eutectic penetration depth. In the results, it is difficult to assume that a brittle layer depth made by eutectic reaction is all of the thickness reduction due to cladding thinning. About 93% cladding thinning of the eutectic penetration depth is favorable as an effective thickness of cladding. And the unreliability of the KALIMER driver fuel pin under the same WPF test condition is lower than that of the WPF pin because of the higher plenum-fuel volume ratio and lower cladding inner radius vs. thickness ratio. KALIMER fuel pin developed from conceptual design has a more stable transient performance for a failure mechanism due to fission gas buildup than the WPF pin.
The Plant-specific Impact of Different Pressurization Rates in the Probabilistic Estimation of Containment Failure Modes
Ahn, Kwang-ll ; Yang, Joon-Eon ; Ha, Jae-Joo ;
Nuclear Engineering and Technology, volume 35, issue 2, 2003, Pages 154~164
The explicit consideration of different pressurization rates in estimating the probabilities of containment failure modes has a profound effect on the confidence of containment performance evaluation that is so critical for risk assessment of nuclear power plants. Except for the sophisticated NUREG-1150 study, many of the recent containment performance analyses (through Level 2 PSAs or IPE back-end analyses) did not take into account an explicit distinction between slow and fast pressurization in their analyses. A careful investigation of both approaches shows that many of the approaches adopted in the recent containment performance analyses exactly correspond to the NUREG-1150 approach for the prediction of containment failure mode probabilities in the presence of fast pressurization. As a result, it was expected that the existing containment performance analysis results would be subjected to greater or less conservatism in light of the ultimate failure mode of the containment. The main purpose of this paper is to assess potential conservatism of a plant-specific containment performance analysis result in light of containment failure mode probabilities.
Development of Movable Detection System for Efficiency Measurement in 3-PM Liquid Scintillation Counting
Hwang, H-Y ; Kwak, S.I ; Cho, Y.H ; Byun, J.I ; Lee, H.Y ; Seo, J.S ; Kwak, J.Y ; Lee, J.M ; Lee, K.B ; Park, T.S ; Chung, K.H ; Lee, C.W ;
Nuclear Engineering and Technology, volume 35, issue 2, 2003, Pages 165~170
We developed an improved 3-PM liquid scintillation counting (3-PM LSC) method in which three detectors can be displaced to back and forth directions, and a data acquisition system being able to provide the values for all parameters required for the method. The detectors are entirely located in a 20-mm lead chamber of an inner diameter of 500 mm. A saw-toothed gear ties up all detectors so as to move them uniformly, up to 50 mm with unit of 1 mm. The data acquisition system was designed in an integrated circuit to perform the necessary works such as fast amplification, discrimination, coincidence and logic analysis. It generates values of nine parameters among twelve's generated in the 3-PH LSC method. The dead time of each counting channel is of extending type, valving from 10 to 100
. We measured the TDCR values with an unquenched liquid scintillation source 1"C by displacing the detectors with a step of 2.5 mm away from counting vial. Their values were derived on the range from 0.9 to 0.6. The extent is three times wider than those regions observed by applying the defocalization technique.ique.
Verification of a Dynamic Compartment Model for the Tritium Behavior in the Plants After Short HTO Release Using a BIOMOVS II Scenario
Park, Heui-Joo ; Kang, Hee-Suk ; Lee, Hansoo ;
Nuclear Engineering and Technology, volume 35, issue 2, 2003, Pages 171~177
A dynamic compartment model was required for the prediction of radiological consequences of the tritiated vapor released from the nuclear facility after an accident. A computer code, ECOREA-T, was developed by incorporating the unit models for the evaluation of tritium behavior in the environment. Dry deposition of tritiated vapor from the atmosphere to the soil was calculated using a deposition velocity. Transport of tritium from the atmosphere to the plant was calculated using a specific activity model, and the result was compared with the Belot's analytic solution. Root uptake of tritiated water from the soil and formation of OBT from T were considered in the model. The ECOREA-T code was verified by comparing the results from the other computer codes using a scenario developed through BIOMOVS II study. The results showed good agreements.