Summary of session W3
Description
This paper describes the Light Water Reactor Thermal/hydraulics Loss-of-Coolant (LOCA) for advanced design Pressurized Water Reactor (PWR) reported LOCA performance of the Westinghouse Advanced 600 MWe PWR (AP600). The AP600 plant features a low power density core, passive safeguards systems and an automatic depressurization system. The Westinghouse BASH and NOTRUMP evaluation models were used for the LOCA analysis. A spectrum of small break LOCA cases (ranging up to an 8 inch diameter line rupture) were assessed and exhibit no core uncovery. The best estimate large break LOCA case computed a peak cladding temperature of 1450 degrees F, well below the regulatory limit of 2200 degrees F. This paper analyzes large break and small break LOCA performance, respectively, for Advanced Light Water Reactor (B-600). The B-600 plant is an 1800 MWt unit that utilizes several design enhancement features including passive safety systems, a large volume pressurizer, and improved elevated coolant loops. A proprietary version of the RELAP/Mod 2 code was used as an evaluation model. Best estimate peak cladding temperature for large break LOCA cases (including several statistical variational cases) was no more than 8200 degrees F. Based on these analyses, it was concluded that licensing criteria can be met with substantial margins. Small break cases were analyzed in this paper, and alternate design concepts for dealing with small break LOCAs were compared. This paper also presents LOCA performance of the Combustion Engineering Advanced Light Water Reactor design (System 80+). The System 80+ is a 3800 MWt plant which incorporates several design enhancements on the proven System 80 design. Small break LOCA analysis (best-estimate procedures) show that the core remains covered for breaks up to 10 inches in diameter. Licensing analysis for large breaks confirm adequate reflooding, acceptable cladding temperature following safety injection system (LPSI)
Additional details
Additional titles
- Subtitle (English)
- Light water reactor thermal/hydraulics
Publishing Information
- Publisher
- American Nuclear Society.
- Imprint Place
- La Grange Park, IL (United States)
- ISBN
- 89448-1460-0
- Imprint Title
- Proceedings of the international topical meeting on safety of next generation power reactors
- Imprint Pagination
- 935 p.
- Journal Page Range
- p. 749.
Conference
- Title
- American Nuclear Society (ANS) topical meeting on the safety of next generation power reactors.
- Dates
- 1-5 May 1988.
- Place
- Seattle, WA (United States).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 23005640
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S99: GENERAL AND MISCELLANEOUS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- DEFORMATION; FUEL ELEMENTS; FUEL-CLADDING INTERACTIONS; LICENSING; LOSS OF COOLANT; PWR TYPE REACTORS; REACTOR CORES; REACTOR SAFETY EXPERIMENTS; SAFEGUARDS; SAFETY STANDARDS; SPECIFICATIONS
- Descriptors DEC
- ACCIDENTS; ENRICHED URANIUM REACTORS; POWER REACTORS; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTORS; STANDARDS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- CONF-880506--.