ROSA-IV LSTF 5% cold leg break analysis using the RELAP5/Mod2 code
Creators
- 1. Paul Scherrer Inst. (PSI), Villigen (Switzerland). Lab. for Thermal-Hydraulics
Description
This paper presents the results of the OECD International Standard Problem calculation no. 26 (ISP-26), performed with the RELAP5/Mod2 code (frozen version 36.06 for the CRAY-X/MP) at the Paul Scherrer Institute (PSI) and analyses their comparison with the 5% cold leg break test (run SB-CL-18) conducted on the Large Scale Test Facility (LSTF) of the ROSA-IV program. The principal objective of the present calculation is the analysis of the simulation capability of the code in regards to the main phenomena occurring during a typical Small Break Loss of Coolant Accident (SB-LOCA), such as: natural circulation heat removal; liquid holdup in the up- and downflow sides of the steam generator U-tubes in combination with a reflux condenser mode heat removal, core upper plenum pressure build-up causing manometric liquid level unbalance between core and downcomer; loop seal clearing concurrent with core liquid depression leading to core heat-up; core level recovery after loop seal clearing; vessel inventory boil-off leading to second core uncovering; core reflooding due to accumulator injection intervention. 7 refs., 15 figs., 2 tabs
Additional details
Publishing Information
- ISBN
- 0-919784-31-3
- Imprint Title
- Canadian Nuclear Society/American Nuclear Society 4. international conference on simulation methods in nuclear engineering
- Imprint Pagination
- 2 v.
- Journal Page Range
- (v.2) [17 p.].
- Report number
- INIS-mf--14622
Conference
- Title
- 4. international conference on simulation methods in nuclear engineering.
- Dates
- 2-4 Jun 1993.
- Place
- Montreal, PQ (Canada).
INIS
- Country of Publication
- Canada
- Country of Input or Organization
- Canada
- INIS RN
- 26073890
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference, Numerical Data
- Descriptors DEI
- EXPERIMENTAL DATA; FLOW MODELS; LOSS OF COOLANT; PWR TYPE REACTORS; R CODES; REACTOR SAFETY EXPERIMENTS; SAFETY ANALYSIS; SIMULATION; TEST FACILITIES; VERIFICATION
- Descriptors DEC
- ACCIDENTS; COMPUTER CODES; DATA; ENRICHED URANIUM REACTORS; INFORMATION; MATHEMATICAL MODELS; NUMERICAL DATA; POWER REACTORS; REACTOR ACCIDENTS; REACTORS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS