• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.435-440
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• 1996

Large Loss-of-Coolant Accidents analyzed in Final Safety Analysis Reports are usually covered by Reactor Inlet Header. Reactor Outlet Header and Primary Pump Suction breaks as representative cases. In this study we analyze the total (guillotine) break of an Emergency Core Cooling System (ECCS) pipe located at the ECCS injection point into the Primary Heat Transport System (PHTS). It was expected that thermal-hydraulic behaviors in the PHT and ECC systems are different from those of a Reactor Inlet Header break, having an equivalent break size. The main purpose of this study is to get insights on the differences occurred between the two cases and to assess these differences from the phenomenon behavior point of view. It was also investigated whether the ECCS line break analysis results could be covered by header break analysis results. The study reveals that as the intact loop has almost the same behavior in both analyzed cases. broken loop behavior is different mostly regarding sheath temperature in the critical core pass and pressure decrease in the broken Reactor Inlet Header. Differences are also met in the ECCS behavior and in event sequences timings.

• 한국마린엔지니어링학회:학술대회논문집
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• 한국마린엔지니어링학회 2001년도 춘계학술대회 논문집
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• pp.106-111
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• 2001

The performance characteristics of heat pump system using the new refrigerant subcooling system were investigated. The new heat pump system has the ice storage tank to accumulate the latent heat of the refrigerant during the night-time. The heat is released to subcool the saturated refrigerant liquid at the outlet of a condenser in the daytime. The experimental apparatus is a well-instrumented heat pump which consisted of a refrigerant loop and a coolant loop. The test sections(condenser and evaporator) were made of tube-in-tube heat exchanger with the horizontal copper tube of 12.7[mm] outer diameter and 9.5[mm] inner diameter. The evaporating temperatures ranged from $-5[^{\circ}C]$ to $0[^{\circ}C]$ and the subcooling degrees of the refrigerant varied from $15[^{\circ}C]$ to $25[^{\circ}C]$. The test of the ice storage was carried out at evaporating temperature of $-10[^{\circ}C]$ and the ice storage mode is an ice-on-coil type. The main results were summarized as follows ; The refrigerant mass flow rate and compressor shaft power of the heat pump system were independent of the subcooling degrees. The cooling capacity o the heat pump system increases as the evaporating temperature and subcooling degree increases. The cooling capacity of the heat pump system is about 25 to 30% higher than that of normal heat pump system. The COP of the heat pump system which subcooled the refrigerant liquid at the outlet of the condenser is about 28% higher than that of the normal heat pump system.

• 한국자동차공학회논문집
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• 제15권3호
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• pp.124-132
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• 2007

The main objective of this work is to investigate the characteristics of a heat pump system for fuel cell electric vehicle (FCEV). The present heat pump system adopts an electrically driven compressor running with R134a and uses the heat from the fuel cell stack as the heat source for the exterior heat exchanger. The experimental work has been done with various operating conditions such as different compressor speeds, fuel cell stack coolant temperatures and flow rates. The heating capacity was measured to be from 4 to 10 kW at $-20^{\circ}C$ ambient temperature, and the outlet temperature of interior heat exchanger was up to $70^{\circ}C$. After 30 seconds from start-up, the system reached a steady state and the heating capacity of 6.8 kW was acquired, and after 90 seconds, the air outlet temperature of interior heat exchanger became $35^{\circ}C$.

• 한국압력기기공학회 논문집
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• 제16권1호
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• pp.11-21
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• 2020

This study aims to develop and advance the evaluation technology for assessing PHWR safety. For this purpose, the complete loss of AC power or station blackout (SBO) was selected as a target accident scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes the main features of the primary heat transport system with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was achieved successfully by running the present model to check out the stable convergence of the key parameters. Subsequently, through the SBO transient analyses two cases with and without the coolant leakage via the PHTS pumps were simulated and the behaviors of the major parameters were compared. The sensitivity analysis on the amount of the coolant leakage by varying its flow area was also performed to investigate the effect on the system responses. It is expected that the results of the present study will contribute to upgrading the evaluation technology of the detailed thermal hydraulic analysis on the SBO transient of the operating PHWRs.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2001년도 추계 학술대회논문집
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• pp.118-125
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• 2001

The performance prediction of SMART MCP was performed using a computational fluid dynamics code. General capacity-head performance curve of MCP, which is provided to other design branches as design input, was obtained and it showed the typical type of axial pump performance curve. When four MCPs operate in parallel and one of them stops while the others continue to operate, SMART requires reduced power operation. A procedure for predicting the performance of SMART MCP for that case was developed and verified with available experimental data. An analysis based on the developed procedure was performed for two cases; the impeller of sloped MCP is fixed or free to rotate in reverse direction. According to the results, $73\%$ flow rate of normal operation enters the reactor core in the case of the locked impeller. In case of the impeller free rotation, the flow rate entering the reactor core is $62.8\%$.

• 한국자동차공학회논문집
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• 제20권4호
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• pp.25-32
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• 2012

Cold start driving cycles exhibit an increases in friction losses due to the low temperatures of metal components and media compared to the normal operating engine conditions. These friction losses are adversely affected to fuel economy. Therefore, in recent years, various techniques for the improvement of fuel economy at cold start driving cycles have been introduced. The main techniques are the upward control of coolant temperature and the fast warm-up techniques. In particular, the fast warm-up techniques are implemented with the coolant flow-controlled water pump and the WHRS (waste heat recovery system). This paper deals with an effect of fast ATF (automatic transmission fluid) warm-up on fuel economy using a recovery system of EGR gas waste heat in a diesel engine. On a conventional diesel engine, two ATF coolers have been connected in series, i.e., an air-cooled ATF cooler is placed in front of the condenser of air conditioning system and a water-cooled one is embedded into the radiator header. However, the new system consists of only a water-cooled heat exchanger that has been changed into the integrated structure with an EGR cooler to have the engine coolant directly from the EGR cooler. The ATF cooler becomes the ATF warmer and cooler, i.e., it plays a role of an ATF warmer if the temperature of ATF is lower than that of coolant, and plays a role of an ATF cooler otherwise. Chassis dynamometer experiments demonstrated the fuel economy improvement of over 2.5% with rapid increase in the ATF temperature.

• 방사성폐기물학회지
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• 제16권3호
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• pp.389-396
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• 2018

국내 가동원전 중 2-루프 가압경수로인 고리1호기는 약 40년 운전한 후, 2017년 6월 18일 영구정지되었다. 영구정지된 고리 1호기는 주요 해체작업을 수행하기전에 계통내 선량률을 저감시켜 작업자피폭을 최소화하기 위한 계통제염을 수행할 예정이다. 일반적으로, 계통제염 범위는 원자로압력용기, 가압기, 증기발생기, 화학 및 체적제어계통, 잔열제거계통 및 원자로 냉각재계통 주요배관을 포함한다. 이러한 계통 및 기기 등을 효율적으로 제염하기 위해서는 제염과정에서 원자로냉각재계통내 유동특성을 평가할 필요가 있다. 계통제염을 위해 순환유량을 제공하는 방법은 다양하나, 본 논문에서는 잔열제거펌프 운전에 따른 고리1호기 원자로냉각재계통내 유동특성을 평가하였다. 잔열제거펌프를 이용한 계통제염은 원자로냉각재 내 유량의 불균형을 초래하여 계통내 기기 및 배관 등에 불순물을 침적시켜 제염이 효율적이지 않다는 것으로 평가되었다.

• 방사성폐기물학회지
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• 제19권1호
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• pp.51-58
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• 2021

Chemical decontamination of primary systems in a nuclear power plant (NPP) prior to commencing the main decommissioning activities is required to reduce radiation exposure during its process. The entire process is repeated until the desired decontamination factor is obtained. To achieve improved decontamination factors over a shorter time with fewer cycles, the appropriate flow characteristics are required. In addition, to prepare an operating procedure that is adaptable to various conditions and situations, the transient analysis results would be required for operator action and system impact assessment. In this study, the flow characteristics in the steady-state and transient conditions for the chemical decontamination operations of the Kori-1 NPP were analyzed and compared via the MARS-KS code simulation. Loss of residual heat removal (RHR) and steam generator tube rupture (SGTR) simulations were conducted for the postulated abnormal events. Loss of RHR results showed the reactor coolant system (RCS) temperature increase, which can damage the reactor coolant pump (RCP)s by its cavitation. The SGTR results indicated a void formation in the RCS interior by the decrease in pressurizer (PZR) pressure, which can cause surface exposure and tripping of the RCPs unless proper actions are taken before the required pressure limit is achieved.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1213-1224
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• 2024

The liquid lead-bismuth cooled fast reactor has been in a single-phase, low-pressure, and high-temperature state for a long time during operation. Considering the requirement of calculation efficiency for long-term transient accident calculation, based on a homogeneous hydrodynamic model, one-dimensional heat conduction model, coolant flow and heat transfer model, neutron kinetics model, coolant and material properties model, this study used the fully implicit difference scheme algorithm of the convection-diffusion term to solve the basic conservation equation, to develop the transient analysis program NUSOL-LMR 2.0 for the lead-bismuth fast reactor system. The steady-state and typical design basis accidents (including reactivity introduction, loss of flow caused by main pump idling, excessive cooling, and plant power outage accidents) for the ABR have been analyzed. The results are compared with the international system analysis software ATHENA. The results indicate that the developed program can stably, accurately, and efficiently predict the transient accident response and safety characteristics of the lead-bismuth fast reactor system.

• 대한전기학회논문지:전기기기및에너지변환시스템부문B
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• 제50권10호
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• pp.483-493
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• 2001

The 3-phase induction is used for the MCP(main coolant pump) and the pure water is used for lubrication of bearing because of the difficulty of repair. Therefore the type of motor is the canned-motor that is welded by sealed can to prevent the stator and rotor from the lubricating water. A lot of Eddy currents are produced in the can because of the conductivity of material. And these eddy currents in the can are the most important cause that decrease the efficiency of induction motor. Therefore we have to find the method to decrease these eddy currents in the can for the improvement of efficiency of motor. In this paper, we proposed the method of design and analysis of canned-motor using equivalent circuit with consideration of can loss for the improvement of efficiency of motor.