• Nuclear Engineering and Technology
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• 제38권1호
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• pp.33-44
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• 2006

This paper presents a review of small-to-medium-sized, pressurized-water-cooled nuclear power reactors whose major primary coolant systems are integrated into a reactor pressure vessel, the concepts categorized as Integrated Primary System Nuclear Reactors (IPSRs). Typical examples of these proposals of interest in this review are CAREM, SMART, IRIS and IMR, all of which are being aimed at the near term deployment. Emphasis is placed on thermal hydraulic aspects. A brief characterization of the IPSR concepts is made and comparisons of plant key parameters are shown. Discussions will follow for the core cooling under rated power conditions and natural circulation heat removal on the basis of the design data available in the public domain.

• International Journal of Fluid Machinery and Systems
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• 제6권1호
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• pp.1-10
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• 2013

Jet pump is one type of artificial lifts and is used when depth and deviation of producing wells increases and pressure depletion occurs. In the present study, numerical analysis has been carried out to analyze the flow behavior and find the performance of the jet pump. Reynolds-averaged Navier Stokes equations were solved and k-${\varepsilon}$ turbulence model was used for simulations. Water and light oil as primary fluids were used to pump water, light oil and heavy oil. The ratios of area and length to diameter of the mixing tube were considered as design parameters. The pump efficiency was considered to maximize for the downhole conditions. It was found that the increase in viscosity and density of the secondary fluid reduced efficiency of the system. Water as primary fluid produced better efficiency than the light oil. It was also found that the longer throat length increased efficiency upto 40% if light oil was used as primary fluid and secondary fluid viscosity was 350 cSt.

• Journal of Mechanical Science and Technology
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• 제17권9호
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• pp.1380-1387
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• 2003

When a cold HPSI (High pressure Safety Injection) fluid associated with an overcooling transient, such as SGTR (Steam Generator Tube Rupture), MSLB (Main Steam Line Break) etc., enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena will arise due to incomplete mixing. If the stratified flow enters the downcomer of the reactor pressure vessel, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. As general thermal-hydraulic system analysis codes cannot properly predict the thermal stratification phenomena, RG 1.154 requires that a detailed thermal-mixing analysis of PTS (pressurized Thermal Shock) evaluation be performed. Also. previous PTS studies have assumed that the thermal stratification phenomena generated in the stagnated loop side of a partially stagnated primary coolant loop are neutralized in the vessel downcomer by the strong flow from the unstagnated loop. On the basis of these reasons, this paper focuses on the development of a 3-dimensional thermal-mixing analysis model using PHOENICS code which can be applied to both partial and total loop stagnated cases. In addition, this paper verifies the fact that, for partial loop stagnated cases, the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is almost neutralized by the strong flow of the unstagnated loop but is not fully eliminated.

• 대한기계학회논문집B
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• 제25권6호
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• pp.837-844
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• 2001

Jet cutting technology currently makes use of a generic supersonic gas jet to improve the cutting speed and performance. In order to get a better understanding of the flow characteristics involved in the supersonic jet cutting technology, the axisymmetric Navier-Stokes equations have been solved using a fully implicit finite volume method. Computations have been conducted to investigate some major characteristics of supersonic coaxial turbulent jets. An assistant gas jet has been imposed on the primary gas jet to simulate realistic jet cutting circumstance. The pressure and the temperature ratios of the primary and assistant gas jets are altered to investigate the major characteristics of the coaxial jets. The total pressure and Mach number distributions, shock wave systems, and the jet core length which characterize the coaxial jet flows are strongly affected by the pressure ratio, but not significantly dependent on the total temperature ratio. The assistant gas jet greatly affects the basic flow characteristics of the shock system and the core length of under and over-expanded jets.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 추계학술대회
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• pp.580-585
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• 2003

The present study addresses experimental results to investigate the details of the near field flow structures produced in the under-expanded, dual, coaxial, swirling, jet. The sonic/supersonic swirling jets are emitted from the sonic inner nozzle and the outer annular nozzle produce the co-swirling and counter swirling against the primary swirling jet, respectively. The interactions between both the secondary annular swirling and primary inner supersonic swirling jets are quantified by the pitot impact and static pressure measurements and visualized by using the Schliern optical method. The experiment is performed for different swirl intensity and pressure ratio. The results obtained show that the secondary co-swirling jet significantly changes the inner under-expanded swirling jet, such as the recirculation zone, pressure distribution, through strong interactions between both the swirling jets and the effects of the secondary counter-swirling jet is similar to the secondary co-swirl jet case.

• 한국압력기기공학회 논문집
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• 제7권4호
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• pp.19-25
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• 2011

Steam Generator(SG) tube is an important component of Nuclear Power Plant(NPP), which is the pressure boundary between the primary and secondary systems. The integrity of SG tube has been confirmed by the eddy current test every outage. The eddy current technique adopting bobbin probe is currently the primary technique for the steam generator tubing integrity assesment. The bobbin probe is one of the essential components which consist of the whole ECT examination system and provides us a decisive data for the evaluation of tube integrity. Until now, all of the ECT bobbin probes in Korea which is necessary to carry out inspection are imported from overseas. However, KHNP has recently developed the bobbin probe design technology and transferred it to domestic manufacturers to fabricate the probes. This study has been conducted to establish technical requirements applicable to the steam generator tube inspection using the bobbin probes fabricated by the domestic manufactures. The results have been compared with the results obtained by using foreign probe to identify the availability to the steam generator tube inspection. As a result, it is confirmed that the domestic bobbin probe is generally applicable to SG tube inspection in the NPPs.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2008년도 제31회 추계학술대회논문집
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• pp.153-156
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• 2008

고공 환경에서의 점화 연소 특성을 확인하기 위해 선행되어야 할 저압/저온 환경 모사 및 연료 유량 제어 실험을 수행하였다. 저압 환경을 모사하기 위하여 초음속 디퓨저를 이용하였고, 공기 유량 공급 및 디퓨저 1차 전단 노즐 압력에 따라 다양한 고도의 저압 환경을 조성할 수 있음을 확인하였다. 또한 저온 환경을 모사하기 위해 액체 질소를 이용한 열교환기를 활용하였고, 혼합 탱크로 유입되는 극저온/상온 공기 온도 조건을 일정하게 유지할 경우 다양한 공기 유량 조건에서 혼합 공기의 온도는 극저온/상온 공기의 혼합비에 의해 결정됨을 알 수 있었다. 이에 따라 본 연구에서 구축한 고고도 환경 모사 시스템을 활용하여 다양한 고도 조건에서의 점화 및 연소 특성 실험 수행이 가능함을 입증하였다.

• Nuclear Engineering and Technology
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• 제42권6호
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• pp.690-703
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• 2010

Sampling based uncertainty analysis was carried out to quantify uncertainty in predictions of best estimate code RELAP5/MOD3.2 for a thermal hydraulic test (10% hot leg break LOCA) performed in the Large Scale Test Facility (LSTF) as a part of an IAEA coordinated research project. The nodalisation of the test facility was qualified for both steady state and transient level by systematically applying the procedures led by uncertainty methodology based on accuracy extrapolation (UMAE); uncertainty analysis was carried out using the Latin hypercube sampling (LHS) method to evaluate uncertainty for ten input parameters. Sixteen output parameters were selected for uncertainty evaluation and uncertainty band between $5^{th}$ and $95^{th}$ percentile of the output parameters were evaluated. It was observed that the uncertainty band for the primary pressure during two phase blowdown is larger than that of the remaining period. Similarly, a larger uncertainty band is observed relating to accumulator injection flow during reflood phase. Importance analysis was also carried out and standard rank regression coefficients were computed to quantify the effect of each individual input parameter on output parameters. It was observed that the break discharge coefficient is the most important uncertain parameter relating to the prediction of all the primary side parameters and that the steam generator (SG) relief pressure setting is the most important parameter in predicting the SG secondary pressure.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.110-110
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• 2002

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 1997년도 봄 학술발표회 논문집
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• pp.147-151
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• 1997

The dynamic characteristics of Control Element Drive Mechanism (CEDM) in Korea standard Nuclear Power Plant was reviewed as a secondary mass in a simplified two degree of freedom system, while the reactor vessel as a primary mass. The design improvement stratege to minimize each displacement amplitude of these primary and secondary masses was proposed. According to this stratege the designs of CEDM components, the shroud and the pressure housing, respectively, were changed using optimization technique.