• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1999년도 추계학술발표회요약집
• /
• pp.267.1-267
• /
• 1999

• Nuclear Engineering and Technology
• /
• 제47권4호
• /
• pp.434-442
• /
• 2015

The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

• 대한전기학회논문지
• /
• 제40권1호
• /
• pp.120-127
• /
• 1991

The purpose of this study is to develop a fault detection and diagnosis scheme that can monitor process fault and instrument fault of a steam generator. The suggested scheme consists of a Kalman filter and two bias estimators. Method of detecting process and instrument fault in a steam generator uses the mean test on the residual sequence of Kalman filter, designed for the unfailed system, to make a fault decision. Once a fault is detected, two bias estimators are driven to estimate the fault and to discriminate process fault and instrument fault. In case of process fault, the fault diagnosis of outlet temperature, feed-water heater and main steam control value is considered. In instrument fault, the fault diagnosis of steam genrator's three instruments is considered. Computer simulation tests show that on-line prompt fault detection and diagnosis can be performed very successfully.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2005년도 추계학술발표회
• /
• pp.416-417
• /
• 2005

• 설비공학논문집
• /
• 제18권7호
• /
• pp.571-577
• /
• 2006

The characteristics of ice packing factor (IPF) at the ice slurry system using one line type are compared with the system using two lines type. The installation space for one transporting line is saved at the one line system. For the one line type, the ice packing factor is reduced along the downstream, but for the two lines type, the ice packing factor is fixed. For the one line system, mass flow rate in the main line is fixed along the down-stream, but for two lines system, the mass flow rate in the main line is reduced along the downstream. For one line system, along the down stream after IPF=0, the temperature at the main steam is increased, and the extracted mass flow is increased. The initial IPF, at which the IPF is not arrived at zero upto the final node, is proposed for the B area.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2003년도 춘계학술발표대회 요약집
• /
• pp.135-135
• /
• 2003

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2002년도 춘계공동학술발표회요약집
• /
• pp.187-187
• /
• 2002

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2003년도 추계학술발표대회 요약집
• /
• pp.199-199
• /
• 2003

• 플랜트 저널
• /
• 제9권3호
• /
• pp.43-47
• /
• 2013

발전 플랜트의 발전량 최적화 및 발전효율 최적화는 에너지공학을 전공한 전문가일지라도 이해하기 힘든 개념이다. 본 연구에서는 엑서지 및 엑서지율이라는 열역학적 상태값을 적용하여 에너지 공학을 전공하지 않은 일반인일지라도 발전량 및 발전효율 최적화 개념을 쉽게 이해할 수 있는 차트가 개발되었다. 발전소의 성능을 파악할 수 있는 대표적인 물성치는 주증기의 온도 및 압력이다. 개발된 차트에서는 주증기의 온도 및 압력에 따른 최대 발전량 곡선과 최대 효율 곡선이 도시되어 있으므로, 해석하고자 하는 발전소의 온도 및 압력을 차트에 적용하여 그 발전소가 얼마만큼 최대 발전량과 최대 효율에 접근해 있는지를 쉽게 파악할 수 있다.

• Nuclear Engineering and Technology
• /
• 제56권9호
• /
• pp.3528-3535
• /
• 2024

SA508 Gr.1A low-alloy steel is being considered as a candidate material for main steam line piping in nuclear power plants. Therefore, improving its strength and toughness is essential for enhancing the leak-before-break (LBB) margin. In this study, six types of model alloys were fabricated by varying the contents of microalloying elements (C, Cu, B, Ti, and Nb) to enhance the mechanical properties of the specimens. The addition of a few ppm of B led to the formation of a fine-grained low-temperature transformation microstructure, resulting in the highest strength among the model alloys. However, the addition of Nb and Ti increased the formation of coarse ferrite, significantly decreasing the strength of the alloys. Reducing the C content while adding a small amount of B simultaneously maintained strength and enhanced toughness. Furthermore, the LBB margins of model alloys and commercial steel were evaluated to validate the influence of varying microalloying content. The model alloys exhibited a substantial increase in yield strength and fracture resistance, resulting in a more than 10% increase in the LBB margin. Notably, the LBB margin of the alloy with 15 ppm B was 1.39, approximately 25% higher than that of commercial steels.