• 한국태양에너지학회 논문집
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• 제28권2호
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• pp.1-9
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• 2008

This paper describes water circulation characteristics of a water/steam receiver at various heat fluxes. The water/steam receiver for a solar tower power system is a natural circulation type. Experimental conditions of water and steam were set at a pressure of 5 bar and temperature of $151.8^{\circ}C$. The experimental device for the water/steam receiver consisted of a steam drum, upper/lower header, riser tubes, and downcomer tube. The experiments were conducted by varying heat fluxes in terms of mass flow rate in each riser tube. However, the total mass flow rate on the riser tubes was fixed at 217.4 g/s. For the uniform heat flux, while the water temperature of the steam drum and upper header were kept at steady state, the temperature of the lower header was fluctuated. For the non-uniform heat flux, while the temperature of the steam drum was kept steady state, the temperature difference increased in the right and left side of the upper header, and the temperature of the lower header was fluctuated.

• 한국소음진동공학회논문집
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• 제15권8호
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• pp.988-995
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• 2005

Vibration has been severly increased at the branch pipe of main steam header since the commercial operation of nuclear power plant. Intense broad band disturbance flow at the discontinuous region such as elbow, valve, and header generates the acoustical pulsation which is propagated through the piping system. The pulsation becomes the source of low frequency vibration at piping system. If it coincide with natural frequency of the pipe system, excessive vibration is made. High level vibration due to the pressure pulsation related to high dynamic stress, and ultimately, to failure probability affects fatally the reliability and confidence of plant piping system. This paper discusses vibration effect for the branch pipe system due to acoustical pulsations by broad band disturbance flow at the large main steam header in 700 MW nuclear power plant. The exciting sources and response of the piping system are investigated by using on-site measurements and analytical approaches. It is identified that excessive vibration is caused by acoustical pulsations of 1.3 Hz, 4.4 Hz and 6.6 Hz transmitted from main steam balance header, which are coincided with fundamental natural frequencies of the piping structure. The energy absorbing restraints with additional stiffness and damping factor were installed to reduce excessive vibration.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2005년도 춘계학술대회논문집
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• pp.780-785
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• 2005

Vibration has been severly increased at the branch pipe of main steam header since the commercial operation of a nuclear power plant. Intense broad band disturbance flow at the discontinuous region such as elbow, valve or heather generates the acoustical pulsation which is propagated through the piping system. The pulsation becomes the source of low frequency vibration at piping system. If it coincide with natural frequency of the pipe system, excessive vibration is made. High level vibration due to the pressure pulsation related to high dynamic stress, and ultimately, to failure probability affects fatally the reliability and confidence of plant piping system. This paper discusses vibration effect for the branch pipe system due to acoustical pulsations by broad band disturbance flow at the large main steam header in 7nn nuclear power plant. The exciting sources and response or the piping system are investigated by using on site measurements and analytical approaches. It is identified that excessive vibration is caused by acoustical pulsations of 1.3Hz, 4.4Hz and 6.6Hz transferred from main steam header, which are coincided with fundamental natural frequencies of the piping structure. The energy absorbing restraints with additional stiffness were installed to reduce excessive vibration.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.495-495
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• 2001

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 추계학술대회
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• pp.373-378
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• 2003

It is important that overheat protection of super heated tube in boiler operation and maintenance. The overheat of super heat tube can make damage and rupture of tube material, which causes accidental shutdown of boiler. The super heated tube overheat is almost due to the lack of uniformity of gas temperature distribution. There are two ways to protect overheat of super heated tube. The one is to control hot gas operation pattern which is temperature or flow distribution. the other is to control super heated steam flow distribution. The former is difficult than the later, because of control device design. In this paper steam flow control method which uses orifices is proposed to protect overheat of super heat tube.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.496-496
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 추계학술발표회요약집
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• pp.348.2-348
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• 2001

• 한국가스학회지
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• 제2권4호
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• pp.1-6
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• 1998

중화학 플랜트의 고온고압 요소인 증기 발생기(steam generator)의 드럼, 헤더 및 헤더 스터브 설계에 응력해석에서 부터 도면자동작도, 제작용 서류 자동생성에 이르기까지의 제반설계과정을 체계적으로 연결시킨 설계정보 통합 관리로 단순 설계오류를 줄이고, 또한 설계변경에 대해 신속한 재설계가 가능한 효율적인 설계를 도모하기 위하여 "피로수명을 고려한 헤더스터브 형상설계 모듈", "TRD301을 기초로한 후육내압부 운전조건(기동/정지 조건) 및 수명평가 모듈", "헤더 및 드럼부 자동작도 모듈"을 개발하였다. 이에 따라 형상설계 모듈을 이용하여 설계된 스터브 형상을 토대로 수명평가 모듈로 수명을 평가한 후 상세설계 도면 및 관련 서류의 자동작도로 이어지는 일관된 종합설계 시스템을 구축하였다.

• 한국전산유체공학회지
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• 제20권1호
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• pp.84-91
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• 2015

SMART adopts, very unique facility, an FMHA to enhance the thermal and flow mixing capability in abnormal conditions of some steam generators or reactor coolant pumps. The FMHA is important for enhancing thermal mixing of the core inlet flow during a transient and even during accidents, and thus it is essential that the thermal mixing characteristics of flow of the FMHA be understood. Investigations for the mixing characteristics of the FMHA had been performed by using experimental and CFD methods in KAERI. In this study, the temperature distribution at the core inlet region is investigated for several abnormal conditions of steam generators using the commercial code, FLUENT 12. Simulations are carried out with two kinds of FMHA shapes, different mesh resolutions, turbulence models, and steam generator conditions. The CFD results show that the temperature deviation at the core inlet reduces greatly for all turbulence models and steam generator conditions tested here, and the effect of mesh refinement on the temperature distribution at the core inlet is negligible. Even though the uniformity of FMHA outlet hole flow increases the thermal mixing, the temperature deviation at the core inlet is within an acceptable range. We numerically confirmed that the FMHA applied in SMART has an excellent mixing capability and all simulation cases tested here satisfies the design requirement for FMHA thermal mixing capability.

• Nuclear Engineering and Technology
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• 제33권4호
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• pp.397-408
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• 2001

A marine reactor shows very different thermal-hydraulic characteristics compared to a land- based reactor. Especially, study on the variation of flow field due to ship motions such as inclination, heaving and rolling is essential since the flow variation has great influence on the reactor cooling capability. In this study, the natural circulation characteristics of integral type marine reactor with modular steam generators were analyzed using computational fluid dynamics code, CFX-4, for inclined conditions. The numerical analyses are performed using the results of natural circulation experiments for integral reactor which are already conducted at Seoul National University. From the results, it was found that the flow rate in the ascending steam generator cassettes increases due to buoyancy effect. Due to this flow variation, temperature difference occurs at the outlets of the each steam generator cassettes. which is mitigated through downcomer by thermal mixing. Also, around the upper pressure header the flow from descending hot leg goes up to the ascending steam generator cassettes due to large natural circulation driving force in ascending steam generator cassettes. From this result, the increase of How rate in the ascending steam generator cassettes could be understood qualitatively.