Attentuation effect of sprinklers on a superheated environment using COMPACT
Description
As a result of a high-energy line break (HELB) in the auxiliary building of a pressurized water reactor, superheated steam is released through a 15.2-cm (6-in.) break in the main stream line. A sophisticated analytical tool is needed to simulate the effects of natural circulation airflow in the building. The COMPACT computer code was used for this study not only because of its ability to model natural circulation airflow but also because it is a recognized and approved tool for outside containment steam-line break applications. Two cases were analyzed in this study. The first case did not model the Appendix R fire protection sprinklers in the break node, while the second case did. Previous conservative analyses have shown that four sprinklers would be available, each providing a flow rate of ∼15 gal/min. A fusible link must fail in order to actuate the sprinklers, and conservative calculations show that the failure time would be 113 s from event initiation. Since the sprinkler headers are already full of water, there is no additional delay time for sprinkler flow to the break compartment
Additional details
Publishing Information
- Journal Title
- Transactions of the American Nuclear Society
- Journal Volume
- 63
- Series
- Trans. Am. Nucl. Soc.
- Journal Page Range
- 255-256
- ISSN
- 0003-018X
- CODEN
- TANSA
Conference
- Title
- Annual meeting of the American Nuclear Society (ANS).
- Dates
- 2-6 Jun 1991.
- Place
- Orlando, FL (United States).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 23036769
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- AIR; C CODES; COMPUTER CODES; COMPUTERIZED SIMULATION; CONTAINMENT SPRAY SYSTEMS; CONTAINMENT SYSTEMS; ENERGY BALANCE; FLOW MODELS; FLOW RATE; HEAT TRANSFER; HYDRAULICS; LOSS OF COOLANT; MASS BALANCE; NATURAL CONVECTION; PERFORMANCE; PRESSURE DEPENDENCE; PRESSURE SUPPRESSION; PWR TYPE REACTORS; REACTOR CORE DISRUPTION; REACTOR SAFETY; STEAM
- Descriptors DEC
- ACCIDENTS; CONTAINMENT; CONVECTION; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; FLUIDS; GASES; MATHEMATICAL MODELS; POWER REACTORS; REACTOR ACCIDENTS; REACTORS; SAFETY; SIMULATION; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- CONF-910603--.