• Nuclear Engineering and Technology
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• 제52권4호
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• pp.755-763
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• 2020

The critical flow phenomenon has been studied because of its significant effect for design basis accidents in nuclear power plants. Transition points from thermal non-equilibrium to equilibrium are different according to the geometric effect on the critical flow. This study evaluates the uncertainty parameters of the critical flow model for analysis of DBA (Design Basis Accident) with the MARS-KS (Multi-dimensional Analysis for Reactor Safety-KINS Standard) code used as an independent regulatory assessment. The uncertainty of the critical flow model is represented by three parameters including the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio, and their ranges are determined using large-scale Marviken test data. The uncertainty range of the thermal non-equilibrium factor is updated by the MCDA (Model Calibration through Data Assimilation) method. The updated uncertainty range is confirmed using an LBLOCA (Large Break Loss of Coolant Accident) experiment in the LOFT (Loss of Fluid Test) facility. The uncertainty ranges are also used to calculate an LBLOCA of the APR (Advanced Power Reactor) 1400 NPP (Nuclear Power Plants), focusing on the effect of the PCT (Peak Cladding Temperature). The results reveal that break flow is strongly dependent on the degree of the thermal non-equilibrium state in a ruptured pipe with a small L/D ratio. Moreover, this study provides the method to handle the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio in the system code.

• Advances in aircraft and spacecraft science
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• 제7권6호
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• pp.537-551
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• 2020

Investigation of leading edge impingement cooling for first stage rotor blades in an aero-engine turbine, its effect on rotor temperature and trailing edge wake loss have been undertaken in this study. The rotor is modeled with the nozzle for attaining a more accurate simulation. The rotor blade is hollowed in order for the coolant to move inside. Also, plenum with the 15 jet nozzles are placed in it. The plenum is fed by compressed fresh air at the rotor hub. Engine operational and real condition is exerted as boundary condition. Rotor is inspected in two states: in existence of cooling technique and non-cooling state. Three-dimensional compressible and steady solutions of RANS equations with SST K-ω turbulent model has been performed for this numerical simulation. The results show that leading edge is one of the most critical regions because of stagnation formation in those areas. Another high temperature region is rotor blade tip for existence of tip leakage in this area and jet impingement cooling can effectively cover these regions. The rotation impact of the jet velocity from hub to tip caused a tendency in coolant streamlines to move toward the rotor blade tip. In addition, by discharging used coolant air from the trailing edge and ejecting it to the turbines main flow by means of the slot in trailing edge, which could reduce the trailing edge wake loss and a total decrease in the blade cooling loss penalty.

• Nuclear Engineering and Technology
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• 제31권1호
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• pp.17-28
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• 1999

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.

• 한국압력기기공학회 논문집
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• 제6권1호
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• pp.43-49
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• 2010

The piping failure probability of the reactor coolant system in Kori unit 1 was evaluated considering stress corrosion cracking. The P-PIE program (Probabilistic Piping Integrity Evaluation Program) developed in this study was used in the analysis. The effect of some variables such as oxygen concentration during start up and steady state operation, and operating temperature, which are related with stress corrosion cracking, on the piping failure probabilities was investigated. The effects of leak detection capability, the size of big leak, piping loops, and reactor types on the piping failure probability were also investigated. The results show that (1) LOCA (loss of coolant accident) probability of Kori unit 1 is extremely low, (2) leak probability is sensitive to oxygen concentration during steady state operation and operating temperature, while not sensitive to the oxygen concentration during start up, and (3) the piping thickness and operating temperature play important roles in the leak probabilities of the cold leg in 4 reactor types having same inner diameter.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3236-3246
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• 2021

Lessons learned from the Fukushima Daiichi nuclear power plant accident directed that multiple failures should be considered more seriously rather than single failure in the licensing bases and safety cases because attempts to take accident management measures could be unsuccessful under the high radiation environment aggravated by multiple failures, such as complete loss of electric power, uncontrollable loss of coolant inventory, failure of essential safety function recovery. In the case of the complete loss of electric power called station blackout (SBO), if there is no mitigation action for recovering safety functions, the reactor core would be overheated, and severe fuel damage could be anticipated due to the failure of the active heat sink. In such a transient condition at CANDU-6 plants, the seal failure of the primary heat transport (PHT) pumps can facilitate a consequent increase in the fuel sheath temperature and eventually lead to degradation of the fuel integrity. Therefore, it is necessary to specify the regulatory guidelines for multiple failures on a licensing basis so that licensees should prepare the accident management measures to prevent or mitigate accident conditions. In order to explore the efficiency of implementing accident management strategies for CANDU-6 plants, this study proposed a realistic accident analysis approach on the SBO transient with multiple-failure sequences such as seal failure of PHT pumps without operator's recovery actions. In this regard, a comparative study for two PHT pump seal failure modes with and without coolant seal leakage was conducted using a best-estimate code to precisely investigate the behaviors of thermal-hydraulic parameters during transient conditions. Moreover, a sensitivity analysis for different PHT pump seal leakage rates was also carried out to examine the effect of leakage rate on the system responses. This study is expected to provide the technical bases to the accident management strategy for unmitigated transient conditions with multiple failures.

• 대한전기학회논문지:전기기기및에너지변환시스템부문B
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• 제49권5호
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• pp.320-326
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• 2000

The canned motor of 3-phase induction is used for main coolant pump(MCP). The type of motor is canned-motor that stator and rotor are welded by sealed can. So, cooling water flows in the air gap of the canned motor as an independent cycling cooling system from the air gap to yoke of the motor to prevent high temperature of stator can and to lubricate bearing. Heat exchange is occurred between cooling water in the air gap and cooling water from the exterior pump to prevent rising of temperature in the motor. I has to analyze the characteristics of can exactly because the loss and the heat in the can are very important to design MCP. Therefore, thermal analysis is studied considering the effect of eddy-current los induced in the can.

• Nuclear Engineering and Technology
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• 제24권3호
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• pp.243-251
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• 1992

소형 냉각재 상실사고시 루프밀봉 형성 및 제거에 대하여 LSTF에서 수행된 실험 SB-CL-18의 결과를 RELAP5/MOD2와 /MOD3를 이용하여 예측하였다. 본 연구는 증기발생기 상향 및 하향 유동에서의 비대칭 냉각재수용에 따른 마노메트릭 유동에 의해 노심노출의 조기발생을 야기시키는 열수력학적 현상을 예측하기 위하여 수행되었다. RELAP5/MOD2를 이용한 해석결과는 루프밀봉 형성 및 제거를 포함하여 감압사고시의 주요 현상을 전반적으로 잘 예측하고 있으나 기초 계산외 결과를 볼 때 현상 및 시간적 순서에 관련하여 몇 가지의 차이가 있었다. RELAP5/MOD3는 RELAP5/MOD2보다 전반적인 현상, 특히 증기발생기 액체수용을 보다 잘 예측하고 있으며, 또 한 RELAP5/MOD3를 이용하여 증기발생기 U자관과 펌프 흡입관의 nodalization수를 늘린 경우는 루프 밀봉제거현상과 시간적 순서를 잘 예측할 수 있었다.

• 한국안전학회지
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• 제8권4호
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• pp.16-20
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• 1993

The combined effect of increased pressure/temperature and the reduced material thicknesses act to increase the stress on the radiator componets. The design life of the radiator is influenced by the cyclic stresses and corrosion, which act to weaken the materials, radiator mechanical failure occurs when a tube or solder Joint ruptures, causing coolant loss or insufficient heat rejection. Therefore, in this study, through strain measurement of the tubes in copper/brass radiator, the strain distribution of the tubes in radiator as function of temperature and pressure is obtained.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3154-3165
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• 2022

During a loss-of-coolant accident (LOCA) in the pressurized water reactor (PWR), there is a possibility that high temperature and internal pressure of the fuel rods lead to ballooning of the cladding, which causes a partial blockage of flow area in a subchannel. Such flow blockage would influence the core coolant flow, thus affecting the core heat transfer during a reflooding phase and subsequent severe accident. However, most of the system analysis codes simulate the accident process based on the assumed channel blockage ratio, resulting in the fact that the simulation results are not consistent with the actual situation. This paper integrates the developed core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module into the self-developed severe accident analysis code ISAA. At the same time, the existing flow blockage model is improved to make it possible to simulate the change of flow distribution due to fuel rod deformation. Finally, the ISAA-FRTMB is used to simulate the QUENCH-LOCA-0 experiment to verify the correctness and effectiveness of the improved flow blockage model, and then the effect of clad ballooning on core heat transfer and subsequent parts of core degradation is analyzed.

• 한국추진공학회지
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• 제9권4호
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• pp.48-54
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• 2005

액체 산소(LOx)와 Jet A-1(Jet engine fuel)을 추진제로 하는 소형 액체 로켓 연소기에서 막 냉각의 효과에 관한 실험적 연구를 수행하였다. 막 냉각제(Jet A-1과 물)는 막냉각장치를 통해 분사되도록 하였다. 막 냉각 유량 변화에 따른 연소기의 외벽온도 및 막 냉각 길이는 추진제 혼합비, 연소실 압력 및 막냉각장치의 형상 변화(분사각)에 따라 비교하였다. 막 냉각에 따른 특성속도 효율의 손실도 막 냉각제를 물과 Jet A-1을 사용하였을 경우에 대해서 각각 구하였다. 연소실 압력의 증가에 따라 노즐에서의 외벽 온도는 증가하였으나, 퍼센트 막냉각 유량이 9% 이상인 경우에 연소실에서는 거의 영향을 받지 않았다. 특성속도는 퍼센트 막냉각 유량이 9% 이상일 때 추진제 혼합비에 영향을 받지 않았다.