Application of fractional scaling methodology (FSM) to loss of coolant accidents (LOCA) Part 3. Process level scaling for peak clad TEMPERATURE
Creators
- 1. UCLA - MAE, Box 951597, 48-121 Engr. IV Los Angeles, CA 90195-1597 (United States)
- 2. Brookhaven National Laboratory Bldg. 474B, Upton, NY 11973 (United States)
- 3. 11 Hamilton Road, Setauket, NY 11733 (United States)
- 4. 703 New Mark Esplanade, Rockville, MD 20850 (United States)
Description
Full text of publication follows: Fractional Scaling Methodology (FSM) is demonstrated at the process level for fuel rod heat up in a nuclear reactor core when the primary system is undergoing a large break loss of coolant accident (LBLOCA.). A large number of test facilities were considered. These facilities vary in size from 1/1700 to 1/21 of a full size PWR. The data for LBLOCA from five integral facilities, namely, PBF, semi-scale, LOBI, THTF and LOFT were analyzed, indicating that peak clad temperature (PCT) occurred in the blowdown phase and was mainly a function of linear heating rate. Similarly, the data for reflood phase from eight integral test facilities, namely, semi-scale, FLECHT, FLECHT-SEASET, FLECHT-SKEWED, PKL, LOFT, CCTF, and SCTF were analyzed indicating that the maximum core heat up occurred in the reflood phase and was mainly function of reflood rate. This approach shows that even with large variations in size and shape, the data can be collapsed and the clad temperature can be predicted by a simple expression of one process variable. The spread in data is indication of other process variables. FSM, when applied on the system, component and process levels, serves to synthesize the wealth of worldwide results from analyses and experiments into compact form for efficient storage, transfer and retrieval of information. This is demonstrated on the process level here. As any general scaling method, FSM can also verify the Phenomena Identification and Ranking Table (PIRT). FSM readily identifies differences in component processes and the effects of scale distortion. (authors)
Availability note (English)
Available in abstract form only, full text entered in this recordAdditional details
Publishing Information
- Imprint Pagination
- 1 p.
- Report number
- INIS-FR--3461
Conference
- Title
- 11. international topical meeting on nuclear reactor thermal hydraulics (Nureth 11)
- Dates
- 2-6 Oct 2005
- Place
- Avignon (France)
INIS
- Country of Publication
- France
- Country of Input or Organization
- France
- INIS RN
- 36045299
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S42: ENGINEERING;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- CORE FLOODING SYSTEMS; DATA COMPILATION; FLOW RATE; HEAT TRANSFER; HEATING RATE; LOSS OF COOLANT; PWR TYPE REACTORS; SCALING LAWS; TEMPERATURE DISTRIBUTION; TEST FACILITIES; THERMAL HYDRAULICS
- Descriptors DEC
- ACCIDENTS; DATA; ECCS; ENERGY TRANSFER; ENGINEERED SAFETY SYSTEMS; ENRICHED URANIUM REACTORS; FLUID MECHANICS; HYDRAULICS; INFORMATION; MECHANICS; POWER REACTORS; REACTOR ACCIDENTS; REACTOR PROTECTION SYSTEMS; REACTORS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS