Boil-off experiments with the EIR-NEPTUN Facility: Analysis and code assessment overview report
Creators
- 1. Paul Scherrer Inst. (PSI), Villigen (Switzerland). Lab. for Thermal-Hydraulics
Description
The NEPTUN data discussed in this report are from core uncovery (boil-off) experiments designed to investigate the mixture level decrease and the heat up of the fuel rod simulators above the mixture level for conditions simulating core boil-off for a nuclear reactor under small break loss-of-coolant accident conditions. The first series of experiments performed in the NEPTUN test facility consisted of ten boil-off (uncovery) and one adiabatic heat-up tests. In these tests three parameters were varied: rod power, system pressure and initial coolant subcooling. The NEPTUN experiments showed that the external surface thermocouples do not cause a significant cooling influence in the rods to which they are attached under boil-off conditions. The reflooding tests performed later on indicated that the external surface thermocouples have some effect during reflooding for NEPTUN electrically heated rod bundle. Peak cladding temperatures are reduced by about 30--40C and quench times occur 20--70 seconds earlier than rods with embedded thermocouples. Additionally, the external surface-thermocouples give readings up to 20 K lower than those obtained with internal surface thermocouples (in the absence of external thermocouples) in the peak cladding temperature zone. Some of the boil-off data obtained from the NEPTUN test facility are used for the assessment of the thermal-hydraulic transient computer codes. These calculations were performed extensively using the frozen version of TRAC-BD1/MOD1 (version 22). A limited number of assessment calculations were done with RELAP5/MOD2 (version 36.02). In this report the main results and conclusions of these calculations are presented with the identification of problem areas in relation to models relevant to boil-off phenomena. On the basis of further analysis and calculations done, changing some of the models such as the bubbly/slug flow interfacial friction correlation which eliminate some of the problems are recommended
Availability note (English)
MF available from INIS under the Report Number; OSTI as TI93011490; NTIS; INIS; GPO.
Files
Additional details
Publishing Information
- Imprint Pagination
- 45 p.
- Report number
- NUREG/IA--0040-3/92
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 24050948
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Descriptors DEI
- CALCULATION METHODS; DATA ACQUISITION SYSTEMS; DRYOUT; FLUID FLOW; FUEL ELEMENT CLUSTERS; HEAT TRANSFER; HYDRAULICS; LOSS OF COOLANT; MEASURING METHODS; NEPTUNE REACTOR; PRESSURE VESSELS; REACTOR COOLING SYSTEMS; STEAM; STEAM SEPARATORS; T CODES
- Descriptors DEC
- ACCIDENTS; COMPUTER CODES; CONTAINERS; COOLING SYSTEMS; ENERGY TRANSFER; EQUIPMENT; EXPERIMENTAL REACTORS; FUEL ASSEMBLIES; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTORS; RESEARCH AND TEST REACTORS; SEPARATION EQUIPMENT; VAPOR SEPARATORS; ZERO POWER REACTORS
Optional Information
- Funding organization
- Nuclear Regulatory Commission, Washington, DC (United States).
- Secondary number(s)
- EIR--629-3/92.