• Nuclear Engineering and Technology
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• 제42권4호
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• pp.394-413
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• 2010

The object of this paper is to identify the key elements that impact a radiation dose at EAB (Exclusion Area Boundary). This study is based on the AST (Alternative Source Terms) as defined in Regulatory Guide 1.183. The LOCA (Loss of Coolant Accident) and the LRA (Locked Rotor Accident) are selected as limiting cases. A sensitivity analysis of accidental behavior with respect to various parameters during LOCA and LRA at Kori Unit 1 is also undertaken for the following objectives: to determine the limiting parameters, to find the impact trend of the radiation dose, and to find the safety margin between AST and TID (Technical Information Document) methodologies. This work confirms that key parameters are particulate removal rate, decontamination factor, iodine chemical form, gap fraction, partitioning factor, and the impact of isotopes group. Comparing TID with AST, the radiation dose of TID is about 80% greater than that of AST under a LOCA, and about 60% greater than that of AST for the case of a LRA; thus the safety margin is remarkably increased when the AST is used. In this work, the sensitivity analysis results are presented in terms of a sensitivity index called the "NDD (Normalized Dose Difference)", which compares the impact of parameters with that of a reference case. These values are derived by using a combination of the leak rate (primary to secondary), iodine chemical form, gap fraction, partitioning factor, spray removal rate, source term, and other variables.

• Journal of Radiation Protection and Research
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• 제42권2호
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• pp.106-113
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• 2017

Background: Interfacing-system loss-of-coolant-accident (ISLOCA) has been identified as the most hazardous accident scenario in the typical PWR plants. The present study as an effort to improve the knowledge of the source term to the environment during ISLOCA focuses on an improvement of the estimation method. Materials and Methods: The improvement was performed to take into account an effect of broken pipeline and auxiliary building structures relevant to ISLOCA. An estimation of the source term to the environment was for the OPR-1000 plants by MELOCR code version 1.8.6. Results and Discussion: The key features of the source term showed that the massive amount of fission products departed from the beginning of core degradation to the vessel breach. Conclusion: The release amount of fission products may be affected by the broken pipeline and the auxiliary building structure associated with release pathway.

• 한국전산구조공학회논문집
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• 제24권3호
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• pp.343-351
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• 2011

원자로 격납건물은 원자력발전소에서 발생가능한 모든 비상사태에 대한 최후의 방벽 역할을 하고 있다. 따라서 사고발생시 원자로 격납건물의 극한능력을 판단하는 것은 매우 중요하다. 대표적인 고려사항 가운데 하나인 LOCA사고 발생시 CANDU형 원자로 격납건물의 극한능력을 파악하기 위해서는 구조적 안전성 평가를 위한 구조해석이 필요하다. CANDU형 원자로 격납건물은 돔과 원통형벽체로 구성된 프리스트레스 콘크리트 쉘 구조물로서 부착식 텐돈을 사용하고 있다. 본 논문에서는 극한내압능력의 평가를 위하여 3차원 구조해석시스템을 사용한 프리스트레스 콘크리트 격납건물의 비선형해석을 수행하였다.

• Nuclear Engineering and Technology
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• 제35권2호
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• pp.91-107
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• 2003

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.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.402-409
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• 2019

When a loss of coolant accident (LOCA) occurs in a nuclear power plant, accident scenarios which can prevent core damage are defined based on break size. Current probabilistic safety assessment evaluates that core damage can be prevented under small-break LOCA (SBLOCA) and steam generator tube rupture (SGTR) with rapid cool down (RCD) strategy when all safety injection systems are unavailable. However, previous research has pointed out a limitation of RCD in terms of initiation time. Therefore, RCD success criteria estimation based on allowable coping time under a SBLOCA or SGTR when all safety injection systems are unavailable was performed based on time-line and thermal-hydraulic analyses. The time line analysis assumed a single emergency operating procedure flow, and the thermal hydraulic analysis utilized MARS-KS code with variables of break size, cooling rate, and operator allowable time. Results show while RCD is possible under SGTR, it is impossible under SBLOCA at the APR1400's current cooling rate limitation of 55 K/hr. A success criteria map for RCD under SBLOCA is suggested without cooling rate limitation.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.1125-1139
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• 2023

The system-integrated modular advanced reactor 100 (SMART100), an integral-type pressurized water small modular reactor, is based on a novel design concept for containment cooling and radioactive material reduction; it is known as the containment pressure and radioactivity suppression system (CPRSS). There is a passive cooling system using a condensation with non-condensable gas in the SMART CPRSS. When a design basis accident such as a small break loss of coolant accident (SBLOCA) occurs, the pressurized low containment area (LCA) of the SMART CPRSS leads to steam condensation in an incontainment refuelling water storage tank (IRWST). Additionally, the steam and non-condensable gas mixture passes through the CPRSS heat exchanger (CHX) submerged in the emergency cooldown tank (ECT) that can partially remove the residual heat. When the steam and non-condensable gas mixture passes through the CHX, the non-condensable gas can interrupt the condensation heat transfer in the CHX and it degrades CHX performance. In this study, condensation heat transfer experiments of steam and non-condensable gas mixture in the natural circulation loop were conducted. The pressure, temperature, and effects of the non-condensable gas were investigated according to the constant inlet steam flow rate with non-condensable gas injections in the loop.

• 기술사
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• 제30권5호
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• pp.82-95
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• 1997

A Study of Improvement in Reactor Thermal Power Measurement Method using Kalman Filter. The objectives of the safety analysis of nuclear power plants are to maintain the surface temperature of fuel and fuel cladding within limit value in case of Loss of Coolant accident (LOCA) so that it ensures the safety and reliability of nuclear power plants. The new technique evaluating the reactor power and improvement of existing plant system increase the safety margin of nuclear power plant operation, and accordingly, economic effect will be anticipated. Hereby, 1 would like to introduce reactor power measurement method using Kalman filter that enables to calculate the reactor power more precisely combining the parameters, for example, turbine output as the 1 st stage pressure of high pressure turbine, and reactor power using energy equilibrium relation. It is expected that the new technique will enhance the accuracy of measurement of reactor power and maintain the reliability of nuclear power operation by increasing operational safety margin, and gain the economic benefit

• 대한설비공학회지:설비저널
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• 제17권4호
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• pp.499-507
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• 1988

The pressure suppression pool of BWR(Boiling Water Reactor) is subjected to hydrodynamic impact in the event of a LOCA(Loss of Coolant Accident). The pressure increase in the reactor dry cell would force the existing water of a vent pipe into the suppression pool. When the water is ejected through the pipe opening into the suppression pool, an abrupt downward force is transmitted to the suppression pool floor. Consequently, many structures installed within the pool must be able to withstand these forces. In order to determine the optimum safe locations of the pool structures, numerical analysis have been carried out to investigate the hydrodynamic behavior of the water jet. In the present analysis, a two-dimensional numerical model is utilized to solve transient flow equations.

• 한국안전학회지
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• 제26권2호
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• pp.6-12
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• 2011

The failure probabilities of pipes in nuclear power plants due to stress corrosion are obtained using the P-PIE program, which is developed for evaluating failure probability of pipes based on the existing PRAISE program. Leak, big leak and LOCA(loss of coolant accident) probabilities are calculated as a function of operating time for several pipes in a domestic nuclear plant. The sensitivity analysis is also performed to find out the important parameters for the failure of pipes due to stress corrosion. The results show that the steady state oxygen concentration and steady state temperature are important parameters and failure probability is very low when the oxygen concentration is maintained according to the regulation.

• Nuclear Engineering and Technology
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• 제55권2호
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• pp.623-639
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• 2023

The specific configuration of the safety systems in VVER-1000/V320 reactors allows a comprehensive study of the Loss of Coolant Accident (LOCA). In the present paper, a verification of the success criteria of the event trees headers for the medium and large break LOCA sequences is conducted. A detailed TRACEV5P5 thermal-hydraulic model of the reactor has been developed, including all safety systems. When analyzing the results of all sequences, some conservatism is observed in certain specific configurations as the success criterion of some headers is not consistent with the classic PSA level 1. Therefore, new proposals for the LOCA event trees are performed based on a reconfiguration of LOCA break ranges and the use of the expanded event trees approach.