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KSME international journal v.17 no.9, 2003년, pp.1380 - 1387  

Thermal-Mixing Analyses for Safety Injection at Partial Loop Stagnation of a Nuclear Power Plant

Hwang, Kyung-Mo   (Department of Structural Integrity &Materials, Korea Power Engineering Company, Inc.CC0007159  ); Kim, Kyung-Hoon   (Department of Mechanical Engineering, College of Advanced Technology, New Clean Powen Plant Technology Research Center, Kyung Hee UniversityUU0001575  );
  • 초록

    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.


  • 주제어

    Thermal Stratification .   HPSI (High pressure Safety Injection) .   Thermal-Mixing Analysis .   Partial Loop Stagnation.  

  • 참고문헌 (9)

    1. Theofanous, T. G. et al., 1984, 'Decay of Buoyancy Driven Stratified Layers with Applications to Pressurized Thermal Shock (PTS) ,' School of Nuclear Engineering and Purdue University, NUREG/CR-3700 
    2. Theofanous, T. G. and Yan, H., 1990, 'A Unified Inter-pretation of One-Fifth to Full Scale Thermal Mixing Experiments Related to Pressurized Thermal Shock,' Division of Systems Research Office of Nuclear Regulatory Commission, NUREG/CR-5677 
    3. RG 1.154, 1987, 'Format and Content of Plant Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors,' Task SI 502-4 
    4. Robinson, H. B., 1985a, 'Pressurized Thermal Shock Evaluation of the H. B. Robinson Unit 2 Nuclear Power Plant,' Oak Ridge National Laboratory, NRC Report NUREG/CR-4183 (ORNL/TM-9567) 
    5. EPRI, 1984, 'Analysis of a Steam Line Break in a Combustion Engineering Pressurized Water Reactor Plant,' NSAC-73 
    6. KEPRI., 1999, 'Pressurized Thermal Shock Evaluation of Kori Unit 1 Reactor Pressure Vessel,' TR. 96NJ12. J1999. 81 
    7. Choi, Y. D. et al., 2000, 'An Experiment on the Flow Control Characteristics of a Passive Fluidic Device,' KSME (B), vol. 24, No.3, pp. 338-345 
    8. CCNPP, 1985b, 'Pressurized Thermal Shock Evaluation of the Calvert Cliffs Unit I Nuclear Power Plant,' Oak Ridge National Laboratory, USNRC Report NUREG/CR-4022 (ORNL/TM-9408) 
    9. CHAM. 1992. 'A Guide to the PHOENICS Input language,' CHAM TR/100 

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