A Proposed Correlation for Critical Flow Rate of Steam/Water Flow
- 1. Thermal Hydraulics Safety Research Division, Korea Atomic Energy Research Institute 111, Daedeok-Daero 989 Beon-Gil, Yuseong-Gu, Daejeon, 34028 (Korea, Republic of)
Description
In the safety analysis of loss-of-coolant-accident (LOCA) and steam generator tube rupture (SGTR) scenarios of light water reactors, a modeling of a break is very important to predict the accident's result. For example, the design of a break simulator to simulate a LOCA in light water reactors requires accurate knowledge of the leakage flow through the break, whose shape can be assumed as an orifice or a nozzle according to the characteristics of the break shape. In the case of a small-break LOCA (SBLOCA) scenario, the critical flow mostly occurs by sub-cooled and saturated liquid water including a two-phase water flow at relatively high-pressure conditions. Among previous critical flow maps, Moody's maps are well known for their critical flow rate and pressure of a single component and two-phase mixture with respect to the upstream enthalpy and pressure. In Moody's model, a slip ratio defined by maximizing the critical flow rate with respect to the slip ratio was introduced, and thus its predictions tended to be too conservative. Later, Moody suggested other maps based on a homogeneous equilibrium model (HEM), and they were useful to predict a choked condition for a given upstream condition. However, their predictions of the critical flow rate seemed to show quite an underestimation compared to various test data. The author suggested more practicable critical flow maps for a critical mass flux and pressure of a steam-water flow with respect to the upstream stagnation conditions using an extended Henry-Fauske model. From a comparison with the selected test data, suggested maps showed conservative predictions within reasonable ranges. In this report, a new correlation for the critical flow rate was developed to simulate the suggested critical flow maps within a reasonable range of error. (authors)
Additional details
Publishing Information
- Journal Title
- Transactions of the American Nuclear Society
- Journal Volume
- 115
- Journal Page Range
- p. 1526-1527
- ISSN
- 0003-018X
Conference
- Title
- 2016 ANS Winter Meeting and Nuclear Technology Expo
- Dates
- 6-10 Nov 2016
- Place
- Las Vegas, NV (United States)
INIS
- Country of Publication
- United States
- Country of Input or Organization
- France
- INIS RN
- 52083266
- Subject category
- S42: ENGINEERING; S22: GENERAL STUDIES OF NUCLEAR REACTORS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- COMPARATIVE EVALUATIONS; COMPUTERIZED SIMULATION; CRITICAL FLOW; CRITICAL MASS; ENTHALPY; FLOW RATE; LIQUIDS; MIXTURES; ORIFICES; SAFETY ANALYSIS; SBLOCA; SIMULATORS; SLIP RATIO; STAGNATION; STEAM GENERATOR TUBE RUPTURE; WATER; WATER COOLED REACTORS; WATER MODERATED REACTORS
- Descriptors DEC
- ACCIDENTS; ANALOG SYSTEMS; DIMENSIONLESS NUMBERS; DISPERSIONS; EVALUATION; FLUID FLOW; FLUIDS; FUNCTIONAL MODELS; HYDROGEN COMPOUNDS; LOSS OF COOLANT; MASS; OPENINGS; OXYGEN COMPOUNDS; PHYSICAL PROPERTIES; REACTOR ACCIDENTS; REACTORS; SIMULATION; THERMODYNAMIC PROPERTIES
Optional Information
- Notes
- 6 refs.; available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US)