• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2008년도 추계학술대회B
• /
• pp.2631-2636
• /
• 2008

The most components and piping of the secondary side of domestic nuclear power plants were manufactured carbon-steel and low-alloy steel. Flow accelerated corrosion leads to wall thinning (metal loss) of carbon steel components and piping exposed to the flowing water or wet steam of high temperature, pressure, and velocity. The feedwater heaters of many nuclear power plants have recently experienced sever wall thinning damage, which increases as operating time progress. Several nuclear power plants in Korea have also experienced wall thinning damage in the shell wall around the impingement baffle. This paper describes the comparisons between the numerical analysis results using the FLUENT code and the experimental results based on down-scaled experimental facility. The experiments were performed based on several types of impingement baffle plates which are installed in low pressure feedwater heater.

• 한국분무공학회지
• /
• 제15권2호
• /
• pp.83-93
• /
• 2010

Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damage, which will increase as operating time progresses. Several nuclear power plants in Korea have experienced wall thinning damage in the area around the impingement baffle inside feedwater heater installed downstream of the turbine extraction stream line. At that point, the extract steam from the turbine is two phase fluid at high temperature, high pressure, and high speed. Since it flows to reverse direction after impinging the impingement baffle, the shell wall of feedwater heaters may be affected by flow-accelerated corrosion. In this paper, to compare degree of shell wall thinning mitigation rate to squared type with mitigation rate of other type baffle plate, four different types of impingement baffle plate-squared, curved, mitigating type and multi-hole type-applied inside the shell. With these comparison data, this paper describes operation of experiments and numerical analysis which is composed similar condition with real feed water heater. And flow visualization is operated for verification of experiments and numerical analysis. In conclusion, this study shows that mitigating type and multi-hole type baffle plate are more effective than other baffle plate about prevention of pressure concentration and pressure value decrease.

• 한국정밀공학회지
• /
• 제27권6호
• /
• pp.55-63
• /
• 2010

Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damage, which will increase as operating time progresses. Several nuclear power plants in Korea have experienced wall thinning damage in the area around the impingement baffle inside feedwater heater installed downstream of the turbine extraction stream line. At that point, the extract steam from the turbine is two phase fluid at high temperature, high pressure, and high speed. Since it flows to reverse direction after impinging the impingement baffle, the shell wall of feedwater heaters may be affected by flow-accelerated corrosion. In this paper, to compare degree of shell wall thinning mitigation rate to squared type with mitigation rate of other type baffle plate, three different types of impingement baffle plate-squared, curved and mitigating type-applied inside the shell. With these comparison data, this paper describes operation of experiments and numerical analysis which is composed similar condition with real feed water heater. And flow visualization is operated for verification of experiments and numerical analysis. In conclusion, this study shows that mitigating type baffle plate is more effective than other baffle plate about prevention of pressure concentration and pressure value decrease.

• 설비공학논문집
• /
• 제19권12호
• /
• pp.821-830
• /
• 2007

Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damage, which will increase as operating time progresses. Several nuclear power plants in Korea have experienced wall thinning damage in the area around the impingement baffle-installed downstream of the high pressure turbine extraction steam line- inside number 5A and 5B feedwater heaters. At that point, the extracted steam from the high pressure turbine is two phase fluid at high temperature, high pressure, and high speed. Since it flows in reverse direction after impinging the impingement baffle, the shell wall of the number 5 high pressure feedwater heater may be affected by flow-accelerated corrosion. This paper describes the comparisons between the numerical results using the FLUENT code and the down scale experimental data on effect of geometry of the impingement baffle plate on the shell wall thinning. Additionally, a new type impingement baffle plate was installed above the impingement baffle plate in the feedwater heater and then the numerical and experimental study were performed in the same progress.

• 대한설비공학회:학술대회논문집
• /
• 대한설비공학회 2009년도 하계학술발표대회 논문집
• /
• pp.847-852
• /
• 2009

Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damange, which will increase as operating time progresses. Several nuclear power plants in Korea have experienced wall thinning damage in the area around the impingement baffle inside feed-water heater installed downstream of the turbine extraction stream line. At that point, the extract steam from the turbine is two phase fluid at high temperature, high pressure, and high speed. Since it flows to reverse direction after impinging the impingement baffle, the shell wall of feedwater heaters may be affected by flow-accelerated corrosion. In this paper, to compare wall thinning range according to change entrance nozzle diameter and position with reference numerical analysis model's wall thinning range, various numerical analysis models applied. In case of changing diameter, four different diameter is applied. And a side of nozzle position, two different position-vertical type and parallel type-is applied. And then this paper describes operation of numerical analysis which is composed similar condition with real feed water heater. In conclusion, this study shows effective design for shall wall thinning by changing nozzle diameter and position.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2004년도 춘계학술대회
• /
• pp.2017-2022
• /
• 2004

There are multistage preheaters in the power generation plan to improve the thermal efficiency of the plant and to prevent the components from the thermal shock. The energy source of these heaters comes from the extracted two phase fluid of working system. These two-phase fluid can cause the so-called Flow Accelerated Corrosion(FAC) in the extracting piping and the bubble plate of the heater for example, in case of point Beach Nuclear Power Plant and in the Wolsung Nuclear Power Plant. The FAC is due to the mass transport of the thin oxide layer by the convection. FAC is dependent on many parameters such as the operation temperature, void fraction, the fluid velocity and pH of fluid and so on. Therefore, in this paper velocity was calculated by FLUENT code in order to find out the root cause of the wall thinning of the feedwater heaters. It also includeed in the fluid mixing analysis model are around the number 5A feedwater heater shell including the extraction pipeline. To identify the relation between the local velocities and wall thinning, the local velocities according to the analysis results were compared with distribution of the shell wall thicknes by ultrasonic test.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1996년도 추계학술발표회논문집(1)
• /
• pp.389-394
• /
• 1996

The conceptual design of Korea Next Generation Reactor (KNGR), which is 3914 MWt PWR, includes the safety depressurization system (SDS) to comply with U.S. NRC's severe accident policy. In this analysis, it is assumed that three Monobloc Sebim valves are adopted for the SDS bleed valves of KNGR. The characteristic of Monobloc Sebim are modeled in the CE-FLASH-4AS/REM code for this analysis. The various feed and bleed (F&B) procedures with Sebim valves are investigated for total loss of feedwater (TLOFW) event. It is found that if operators open two out of three Sebim valves in conjunction with four HPSI pumps before hot leg temperature reaches saturation condition, the decay heat removal and core inventory make-up function can be successfully accomplished. Therefore, this F&B procedure can be used for mitigating the TLOFW event of the KNGR. This result also demonstrates the feasibility of adopting the Monobloc Sebim valves for the SDS of KNGR.

• 에너지공학
• /
• 제23권1호
• /
• pp.90-94
• /
• 2014

본 논문은 원자력발전소의 내환경기기검증(EQ)을 위한 HVAC 과도분석 방법에 대한 내용을 기술하고 있다. 분석 대상 격실은 비안전관련 HVAC 계통에 의해 공급되는 격실 중에 원자로 안전정지를 담당하는 중요기기가 위치한 구역/격실을 선정하였다. 그리고 해당 HVAC 계통이 공급되는 건물별로 HVAC 과도시 온도조건을 분석하였다. 본 분석을 위해서 GOTHIC 전산코드를 사용하였다. 온도분석 결과는 원자로 보조건물 환기계통(DVN)의 W315/W415 격실에서 $82.2^{\circ}C$로 가장 높은 온도값을 나타내며, 제어봉구동장치 전원공급건물 및 보조급수펌프실(DVG) 계통의 W229 (Auxiliary feedwater pump room) 격실에서 $52.7^{\circ}C$, 기기냉각건물 환기계통(DVI)의 전 격실에서 $42.9^{\circ}C$, 전기건물 주환기 계통(DVL)의 L207 (Hot workshop) 격실에서 $57.3^{\circ}C$를 각각 나타났다. 이러한 온도값은 일반적인 원전의 기기검증 제한값인 $171^{\circ}C$이하이므로 내환경검증 요건을 만족하는 온도이다.

• Nuclear Engineering and Technology
• /
• 제41권3호
• /
• pp.287-294
• /
• 2009

Due to thermal hydraulics phenomena, such as thermal stratification, various events occur to the parts of a nuclear power plant during their lifetimes: e.g. cracked and dislocated pipes and thermally fatigued, bent, and damaged supports. Due to the operational characteristics of the parts of the steam generator feedwater inlet horizontal pipe, thermal stratification takes place particularly frequently. However, the thermal stress due to thermal stratification at the steam generator feedwater inlet horizontal pipe was not reflected in the design stage of old plants(Kori Unit No.1, 2, 3 and 4, Yeonggwang Unit No. 1 and 2, and Uljin Unit No. 1 and 2; referred to as old-style power plants hereinafter). Accordingly, a verification experiment was performed for thermal stratification in the horizontal inlet nozzle steam generator of old-style plants. If thermal stratification occurred in the horizontal pipe of an old-style power plant, numerical analysis of the temperature distribution of the pipes and fluids was conducted. The temperature distributions were compared at the curved part of the pipe and the horizontal pipe before and after the installation of the improved thermal sleeves designed to alleviate thermal stress due to thermal stratification. The thermal stress reduction measure was proven effective at the steam generator inlet horizontal pipe and the curved part of the pipe.

• 에너지공학
• /
• 제28권3호
• /
• pp.1-9
• /
• 2019

화력발전소에서 사용되는 급수 가열기 튜브에서는 사용중에 두께 감육이 발생하여 수명이 소진된다. 감육에 의한 파열 우려가 있으면 수명이 종료되는데, 파열조건을 결정하는 튜브 벽의 응력은 내압에 의한 원주방향 응력의 영향이 가장 큰 것으로 알려져 있지만, 튜브 내외부 온도차이에 의한 열응력에 대한 고려 또한 필요하다. 튜브 두께 방향의 온도차이는 열응력을 발생시켜 튜브의 잔여수명을 단축시키는 영향을 준다. 본 논문에서는 급수가열기 내에서 튜브 내표면과 외표면에 온도 차이가 가장 큰 과열저감구역(de-superheating zone)을 대상으로 열응력을 연구하였다. 원주방향으로 균일하게 감육된 튜브에서 두께방향의 온도차 때문에 발생하는 원주방향 응력, 반경방향 응력 및 온도분포를 평가하기 위한 해석적 수식을 제시하였다. 제시된 해석식의 정확도와 효과를 검증하기 위해 식으로부터의 계산된 결과를 유한요소해석으로 평가한 정확한 결과와 비교하였다. 또한, 유한요소해석으로 편심 감육된 튜브에 대한 응력도 평가하였다. 열응력 해석 및 온도 분포 해석에서 대류열전달 계수의 영향을 분석하기 위해 튜브 내표면 및 외표면에 여러 값의 열대류 계수를 적용하여 해석 결과를 비교하였다. 해석 결과 튜브 내표면보다 외표면의 열대류 계수가 응력 발생에 더 큰 영향을 주는 것으로 나타났다. 열하중만 고려된 경우, 균일 감육과 편심 감육 상태 모두에서 원주방향 응력이 반경방향 응력보다 크게 평가되었다.