Pressure loadings of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) reactor release mitigation structures from large-break LOCAs
- Argonne National Lab., IL (USA)
Analyses have been carried out of the pressurization of the accident release mitigation structures of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) pressurized water reactors following large-break loss-of-coolant accidents. Specific VVER systems for which calculations were performed are the VVER-440 model V230, VVER-440 model V213, and VVER-1000 model V320. Descriptions of the designs of these and other VVER models are contained in the report DOE/NE-0084. The principal objective of the current analyses is to calculate the time dependent pressure loadings inside the accident localization or containment structures immediately following the double-ended guillotine rupture of a primary coolant pipe. In addition, the pressures are compared with the results of calculations of the response of the structures to overpressure. Primary coolant system thermal hydraulic conditions and the fluid conditions at the break location were calculated with the RETRAN-02 Mod2 computer code (Agee, 1984). Pressures and temperatures inside the building accident release mitigation structures were obtained from the PACER (Pressurization Accompanying Coolant Escape from Ruptures) multicompartment containment analysis code developed at Argonne National Laboratory. The analyses were carried out using best estimate models and conditions rather than conservative, bounding-type assumptions. In particular, condensation upon structure and equipment was calculated using correlations based upon analyses of the HDR, Marviken, and Battelle Frankfurt containment loading experiments. The intercompartment flow rates incorporate an effective discharge coefficient and liquid droplet carryover fraction given by expressions of Schwan determined from analyses of the Battelle Frankfurt and Marviken tests. 5 refs., 4 figs.
- Research Organization:
- Argonne National Lab., IL (USA)
- Sponsoring Organization:
- DOE/NE
- DOE Contract Number:
- W-31109-ENG-38
- OSTI ID:
- 5530200
- Report Number(s):
- CONF-890855-51; ON: DE89017697; TRN: 89-027118
- Resource Relation:
- Conference: 10. international conference on Structural Mechanics in Reactor Technology (SMIRT), Anaheim, CA (USA), 14-18 Aug 1989
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LOSS OF COOLANT
MITIGATION
PWR TYPE REACTORS
REACTOR COOLING SYSTEMS
CONTAINMENT
D CODES
FISSION PRODUCT RELEASE
HEAT TRANSFER
HYDRAULICS
PRESSURIZING
R CODES
REACTOR SAFETY
TRANSIENTS
ACCIDENTS
COMPUTER CODES
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
FLUID MECHANICS
MECHANICS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SAFETY
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled