Two-phase 3 x 3 rod bundle test facility for post-critical heat flux boiling
Description
This report describes the rod bundle post-CHF tests in progress and the test facility at Lehigh University. The mechanical and electrical design of the experimental facility and the iterative process used to arrive at the choices made for the design are described in detail. The test facility consists of a nine (3 x 3) rod bundle in a square shroud which form the test section together with the hot patches at the top and bottom ends. The rods and the hot patches are electrically heated while the shroud is radiatively heated. The test section includes instrumentation to measure the vapor superheat temperature and pressure drop upstream and downstream of a rod gap spacer. This is the first application of the hot patch technique for generating post-CHF conditions in a rod bundle and thus quasi-steady-state tests are being thought of as a backup procedure for conducting these post-CHF heat transfer tests. The test section is part of a well instrumented recirculating loop to generate the desired post-CHF conditions. The other major components of the heat transfer loop include the surge tank, pumps, boiler, separation tank and condenser
Availability note (English)
MF available from INIS under the Report Number; Available from NTIS, PC A04/MF A01 - GPO* $4.25 as TI84901536.
Files
Additional details
Additional titles
- Augmented title (English)
- PWR; BWR
Publishing Information
- Imprint Pagination
- 56 p.
- Report number
- NUREG/CR--3849
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 16010919
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Descriptors DEI
- BWR TYPE REACTORS; CRITICAL HEAT FLUX; FUEL ASSEMBLIES; HEAT TRANSFER; HYDRAULICS; LOSS OF COOLANT; PWR TYPE REACTORS; REACTOR SAFETY; TESTING; TWO-PHASE FLOW
- Descriptors DEC
- ACCIDENTS; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; FLUID FLOW; HEAT FLUX; POWER REACTORS; REACTOR ACCIDENTS; REACTORS; SAFETY; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS