Pressure Drop-Flow Rate Curves for Single Phase Steam in Steam Generator U-tubes During Severe Accidents
- Korea Atomic Energy Research Institute: Daedeok-Daero 989-111, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)
Thermally induced steam generator tube rupture (TI-SGTR) can occur in pressurized water reactors (PWRs) with U-tube steam generators (SGs) during high pressure severe accident sequences resulting in containment bypass. Robust natural circulation patterns including in-vessel natural circulation and countercurrent flow to and from the SGs of superheated steam and other hot gases heats cooler structures in the reactor coolant system (RCS). The buoyancy driven flows return cooler and denser gases back to the core slowing the progression of some core degradation phenomena and failure of the lower head. As much as 50 percent of the decay energy and energy produced from core degradation oxidation reactions can be transported from the core to SGs resulting in the heat up and thermal creep of RCS pressure boundary components including hot leg nozzles, pressurizer surge line, and the SG tubes, many of which contain flaws. The creep failure of an individual tube is dependent on the transitory heating of the tube, a function of the mass flow rate through the tube and local temperature history of steam entering the tube. Buoyancy driven natural circulation flow rates are determined by changes in gravitational head, a combination of changes in fluid density due to heat transfer and change in elevation, and irreversible pressure drops resulting primarily from friction loss in the flow paths and other form losses. The nonuniform flow distribution in U-tubes of varying lengths and heights during single-phase liquid and two-phase natural circulation has been extensively studied in scaled test facilities and verified numerically in previous works. We hypothesize the countercurrent natural circulation of single-phase superheated steam in U-tube SGs is governed by the same characteristic pressure drop-flow rates curves of single-phase liquid and two-phase natural circulations in U-tubes connected to common plena derived in. Secondly, we propose flow distribution is a function of U-tube geometry - height, length, hydraulic diameter, etc., and the larger SGs of Combustion Engineering (CE) style 2 x 4 PWRs, which are the focus of this study, will show greater nonuniformity in tube by tube flow rates and ratio of forward flowing tubes to reverse flowing tubes. This work derives characteristic pressure drop-flow rate curves for single-phase superheated steam in CE-type SGs during a high pressure station blackout (SBO) sequence with failed auxiliary feedwater (AFW) and dry secondary side which is closely related to the TI-SGTR event. (authors)
- OSTI ID:
- 23042957
- Journal Information:
- Transactions of the American Nuclear Society, Vol. 115; Conference: 2016 ANS Winter Meeting and Nuclear Technology Expo, Las Vegas, NV (United States), 6-10 Nov 2016; Other Information: Country of input: France; 2 refs.; available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US); ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
Similar Records
Assessment of RELAP5/MOD3.1 with the LSTF SB-SG-06 experiment simulating a steam generator tube rupture transient
MELCOR Analysis of Steam Generator Tube Creep Rupture in Station Blackout Severe Accident
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FEEDWATER
FLOW RATE
FRICTION
HEAT EXCHANGERS
HEATING
HYDRAULICS
NATURAL CONVECTION
PRESSURE DROP
PRESSURIZERS
PWR TYPE REACTORS
REACTOR COOLING SYSTEMS
SEVERE ACCIDENTS
STATION BLACKOUT
STEAM GENERATOR TUBE RUPTURE
STEAM GENERATORS
TEST FACILITIES
BUOYANCY