Analyses of corium spreading in Mark I containment geometry
Description
An assessment of melt spreading in the Mark I system has been carried out using the MELTSPREAD-1 computer code together with supporting analyses. Application of MELTSPREAD-1 confirms the calculation of shell survival in a wet containment for the most probable melt release conditions from NUREG/CR-5423. According to MELTSPREAD-1, a dry containment also may not be threatened by melt spreading. This reflects the heat losses undergone by the melt in the process of spreading to the shell conservatively neglected in NUREG/CR-5423. However, there exist parameter ranges outside the most probable set where shell failure may be calculated. Accounting for the breakup and quenching of melt relocating through a deep layer of subcooled water also conservatively neglected in NUREG/CR-5423 can reduce the set of parameter variations for which containment failure is calculated in the wet case
Additional details
Publishing Information
- Imprint Title
- Proceedings of the US Nuclear Regulatory Commission nineteenth water reactor safety information meeting. Volume 2, Severe accident research; Severe accident policy implementation; Accident management
- Imprint Pagination
- 504 p.
- Journal Page Range
- p. 265-290.
- Report number
- NUREG/CP--0119-Vol.2
Conference
- Title
- 19. Nuclear Regulatory Commission (NRC) water reactor safety information meeting.
- Dates
- 28-30 Oct 1991.
- Place
- Bethesda, MD (United States).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 24027127
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- A CODES; AFTER-HEAT; BWR TYPE REACTORS; COMPUTER CODES; COMPUTERIZED SIMULATION; CONCRETES; CONTAINMENT SYSTEMS; COOLING; CORIUM; FAILURES; FLOW RATE; HEAT TRANSFER; HYDRAULICS; LINERS; LOSS OF COOLANT; M CODES; MELTDOWN; PEACH BOTTOM-2 REACTOR; PEACH BOTTOM-3 REACTOR; PRESSURE VESSELS; PROBABILITY; PWR TYPE REACTORS; QUENCHING; RADIOACTIVITY TRANSPORT; REACTOR ACCIDENTS; REACTOR CORE DISRUPTION; REACTOR SAFETY; RESEARCH PROGRAMS; SENSITIVITY ANALYSIS; SOLIDIFICATION; THREE MILE ISLAND-2 REACTOR; US NRC
- Descriptors DEC
- ACCIDENTS; BUILDING MATERIALS; CONTAINERS; CONTAINMENT; ENERGY TRANSFER; ENGINEERED SAFETY SYSTEMS; ENRICHED URANIUM REACTORS; HEAT TREATMENTS; MATERIALS; NATIONAL ORGANIZATIONS; POWER REACTORS; REACTORS; SAFETY; SIMULATION; THERMAL REACTORS; US ORGANIZATIONS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- CONF-911079--Vol.2.