Flooding experiments with steam and water in a large diameter vertical tube
- 1. Department of Nuclear Engineering, Texas A and M University, TX (United States)
Description
An experimental study on flooding in a large diameter tube is being conducted. In a countercurrent, two-phase flow system, flooding can be defined as the onset of flow reversal of the liquid component which results in cocurrent flow. Flooding can be perceived as a limit to two-phase countercurrent flow, meaning that pairs of liquid and gas flow rates exist that define the envelope for stable countercurrent flow for a given system. Flooding in the AP600 pressurizer surge line can affect the vessel refill rate following a small break LOCA. Analysis of hypothetical severe accidents with current simplified flooding models show that these models represent the largest uncertainty in steam generator tube creep rupture. During a hypothetical station blackout scenario without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. Experiments have been conducted in a 3-inch (76.2 mm) diameter tube with subcooled water and superheated steam as the working fluids at atmospheric pressure. Water flows down the inside of the tube as an annulus while the steam flows upward in the middle. Water flow rates vary from 3.5 to 12 GPM (0.00022 to 0.00076 m3/s) and the water inlet temperature is about 70degC. The steam inlet temperature is about 110degC. It was found that a larger steam flow rate was needed to achieve flooding for a lower water flow rate and for a higher water flow rate. This unique data for flooding in steam-water systems in large diameter tubes will reduce uncertainty in flooding models currently utilized in reactor safety codes. (author)
Additional details
Publishing Information
- Imprint Title
- NURETH-13: Proceedings of the 13th international topical meeting on nuclear reactor thermal hydraulics
- Imprint Pagination
- [4617 p.]
- Journal Page Range
- [16 p.]
Conference
- Title
- 13. international topical meeting on nuclear reactor thermal hydraulics
- Acronym
- NURETH-13
- Dates
- 27 Sep - 2 Oct 2009
- Place
- Kanazawa, Ishikawa (Japan)
INIS
- Country of Publication
- Japan
- Country of Input or Organization
- Japan
- INIS RN
- 43102464
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- CREEP; FORECASTING; HYPOTHETICAL ACCIDENTS; LOSS OF COOLANT; PRESSURIZERS; PWR TYPE REACTORS; REACTOR SAFETY; RUPTURES; SAFETY ANALYSIS; STEAM; STEAM GENERATORS; SURGES; THERMAL HYDRAULICS; TUBES; TWO-PHASE FLOW
- Descriptors DEC
- ACCIDENTS; BOILERS; ENRICHED URANIUM REACTORS; FAILURES; FLUID FLOW; FLUID MECHANICS; HYDRAULICS; MECHANICAL PROPERTIES; MECHANICS; POWER REACTORS; REACTOR ACCIDENTS; REACTORS; SAFETY; THERMAL REACTORS; VAPOR GENERATORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- Available as CD-ROM Data in PDF format, Folder Name: FullPaper, Paper ID: N13P1094.pdf; 19 refs., 8 figs., 1 tab.