Analysis of an AP600 intermediate-size loss-of-coolant accident
Description
A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations preformed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important
Additional details
Publishing Information
- Imprint Title
- Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 2, Sessions 6-11
- Imprint Pagination
- 795 p.
- Journal Page Range
- p. 1153-1168.
- Report number
- NUREG/CP--0142-Vol.2
Conference
- Title
- 7. international meeting on nuclear reactor thermal hydraulics.
- Dates
- 10-15 Sep 1995.
- Place
- Saratoga, NY (United States).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 27019787
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference, Numerical Data
- Descriptors DEI
- DIAGRAMS; EXPERIMENTAL DATA; HYDRAULICS; LOSS OF COOLANT; N CODES; PWR TYPE REACTORS; R CODES; REACTOR SAFETY; SCALE MODELS; T CODES; THEORETICAL DATA
- Descriptors DEC
- ACCIDENTS; COMPUTER CODES; DATA; ENRICHED URANIUM REACTORS; INFORMATION; NUMERICAL DATA; POWER REACTORS; REACTOR ACCIDENTS; REACTORS; SAFETY; STRUCTURAL MODELS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- CONF-950904--Vol.2.