• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3919-3927
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• 2022

This study presents an optimal design of the coolant system of a non-refueling full-life small reactor by analyzing the space-integrated geometrical and electromagnetic variables of an extra vessel electromagnetic pump (EVEMP) for the circulation of a lead-bismuth eutectic (LBE) coolant. The EVEMP is an ideal alternative to the thermal-hydraulic system of non-refueling full-life micro reactors as it possesses no internal structures, such as impellors or sealing structures, for the transportation of LBE. Typically, the LBE passes through the annular flow channel of a reactor, is cooled by the heat exchanger, and then circulates back to the EVEMP flow channel. This thermal-hydraulic flow method is similar to natural circulation, which enhances thermal efficiency, while providing a golden time for cooling cores in the event of an emergency. When the forced circulation technology of the EVEMP was applied, the non-refueling full-life micro reactor achieve an output power of 60 MWt, which is higher than that achievable via the natural circulation method (30 MWt). Accordingly, an optimized EVEMP for Micro URANUS with a flow rate of 4196 kg/s and developed pressure of 73 kPa under a working temperature of 250 ℃ was designed.

• Journal of Advanced Marine Engineering and Technology
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• 제31권1호
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• pp.84-94
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• 2007

As a reactor coolant pump(RCP) is operated in the nuclear power system for a long time. so its surface is continuously contaminated by radioactive scales. In order to perform regular or emergency repair about RCP internals a special decontamination process should be used to reduce the radiation from the RCP surface by means of chemical cleaning. In this study, applicable possibility in chemical decontamination for RCP was investigated on the various materials. The STS 304 showed the best electrochemical properties for corrosion resistance than other materials. However, the pitting corrosion was slightly generated in both STS 415 and STS 431 with the increasing numbers of cycle and intergranular corrosion were sporadically observed. The size of their pitting corrosion and intergranular corrosion were also increased with increasing cycle numbers.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.721-722
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• 2005

• 대한전기학회:학술대회논문집
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• 대한전기학회 2011년도 제42회 하계학술대회
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• pp.1760-1761
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• 2011

원자력발전기술원은 발전정지 관련계통 제어케비넷 내에 장착된 제어용 기기들의 다중화 설계변경 활동을 지원하고 관련 기기의 배선상태 등의 육안점검을 통해 취약성 여부를 최종 확인하기 위하여 국내 Westinghouse형 원전 계측제어 케비넷 점검을 수행하였다. 또한 관련 설계변경에 대한 신뢰도평가 기술지원도 함께 수행하여 해당 설계변경이 설비의 신뢰도 향상에 효과가 있는지를 정량적으로 평가하고자 하였다. 이에 따라 원자로냉각재펌프(RCP, Reactor Coolant Pump) 제어 채널의 다중화 개선에 대하여 설계변경 전후의 기기 배열 변화에 따른 계통 신뢰도 변화를 대표유형 기기의 고장률에 근거하여 분석하였다. 고장수목을 이용하여 설계변경 전후의 RCP 고장정지로 인한 발전정지를 유발하는 고장조합을 도출하고, 고장정지 확률 변화를 정량화 하였다. 또한 기기 보호 측면에서 펌프 보호를 위한 신호를 출력하지 못하는 경우를 정량화하여 이를 비교하였다.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.368-378
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• 2018

Understanding of turbulent flow in the reactor coolant pump (RCP) is a premise of the optimal design of the RCP. Flow structures in the RCP, in view of the specially devised spherical casing, are more complicated than those associated with conventional pumps. Hitherto, knowledge of the flow characteristics of the RCP has been far from sufficient. Research into the nonintrusive measurement of the internal flow of the RCP has rarely been reported. In the present study, flow measurement using particle image velocimetry is implemented to reveal flow features of the RCP model. Velocity and vorticity distributions in the diffuser and spherical casing are obtained. The results illuminate the complexity of the flows in the RCP. Near the lower end of the discharge nozzle, three-dimensional swirling flows and flow separation are evident. In the diffuser, the imparity of the velocity profile with respect to different axial cross sections is verified, and the velocity increases gradually from the shroud to the hub. In the casing, velocity distribution is nonuniform over the circumferential direction. Vortices shed consistently from the diffuser blade trailing edge. The experimental results lend sound support for the optimal design of the RCP and provide validation of relevant numerical algorithms.

• 한국압력기기공학회 논문집
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• 제19권1호
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• pp.44-51
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• 2023

The RCP is one of the main components in nuclear power plants and plays an important role in circulating coolant to the RCS system. Currently, nuclear plants are monitored using various monitoring systems. However, since they operate independently according to their functional purpose, it is not able to analyze vibration and operation/performance information comprehensively, and thus failure diagnosis accuracy is limited. In addition, these systems do not provide some important information (such as fault type, parts and cause) necessary for emergency actions, but provide only alarm information. To improve these technical problems, this study proposes a diagnosis technique (M/L, Rule-based model, Data-driven model, Narrow band model) and methodology for comprehensive analysis.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2011년도 추계학술대회 논문집
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• pp.221-226
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• 2011

The structural integrity of APR1400 reactor vessel internals has been being assessed referring the US Nuclear Regulatory Commission regulatory guide 1.20 comprehensive vibration assessment program. The program is composed of a vibration and stress analysis, a limited vibration measurement, and an inspection. This paper covers the vibration and stress analysis on the reactor vessel internals due to the pulsation of reactor coolant pump. 3-dimensional models to calculate the hydraulic loads and structural responses were built and the pressure distributions and the structural responses were predicted using ANSYS. The peak stress of the reactor vessel internals is much lower than the acceptance limit.

• 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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 추계학술발표회요약집
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• pp.254.2-254
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.132-132
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• 2002