Design, Feasibility, and Testing of Instrumented Rod Bundles to Improve Heat Transfer Knowledge in PWR Fuel Assemblies

Conference · Sun Jul 01 04:00:00 EDT 2007

OSTI ID:21229263

Bergeron, A ^[1]; Chataing, T; Garnier, J ^[2]; Decossin, E; Peturaud, P ^[3]; Yagnik, S K ^[4]

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1. CEA, Saclay (France)
2. CEA, Genoble (France)
3. EDF/R and D, Chatou (France)
4. Electric Power Research Institute - EPRI (United States)

Two 5 x 5 test rod bundles mimicking the PWR fuel assembly have been adapted into two suitable test loop facilities, respectively, to carry out sufficiently detailed hydraulic and thermal measurements in identical geometric configuration. The objective is to investigate heat transfer phenomena in single-phase as well as with onset of nucleate boiling (ONB). The accuracy and reproducibility of the temperature measurements using the sliding-traversing thermocouple device under typical PWR conditions has been demonstrated in the thermal test facility. In the hydraulic loop, a Laser Doppler Velocimetry (LDV) system to precisely scan the local axial velocity component in each sub-channel has been implemented. The approach is to utilize mean sub-channel axial velocity distributions and pressure drop data from the hydraulic loop and the global boundary conditions (Pressure, Temperature, flow rate) from the thermal loop to simulate sub-channels in appropriate T/H codes. This permits computation of sub-channel averaged fluid temperatures (as well as mass velocity) in various subchannels within the test bundle. Subsequently, in conjunction with the wall temperatures and applied heat flux values from the thermal loop, it is possible to develop a complete map of heat transfer coefficients along the 9 instrumented central heater rods. Locations downstream of spacer grids would be of special interest. Depending on pressure, mass velocity and heat flux conditions of a given test, the inlet temperature will be a parameter to be varied so that the ONB boundary can be observed within the bundle. Detailed designs of the test section, required loop modifications, and adaptation of specialized instrumentation and data acquisition systems have been accomplished in both test loops. Further we have established that based on such detailed rod surface temperature and sub-channel axial velocity measurements, it is possible to achieve sufficient accuracy in the temperature measurements to meet the objective of improving the heat transfer correlations applicable to PWR cores. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 21229263
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
42 ENGINEERING
BOUNDARY CONDITIONS
CALCULATION METHODS
DATA ACQUISITION SYSTEMS
DESIGN
FLOW RATE
FUEL ELEMENT CLUSTERS
H CODES
HEAT FLUX
HEAT TRANSFER
NUCLEATE BOILING
PRESSURE DROP
PWR TYPE REACTORS
T CODES
TEMPERATURE MEASUREMENT
TEST FACILITIES
TESTING
THERMOCOUPLES
VELOCITY