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Title: Nucleate boiling pressure drop in an annulus: Book 5

The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. Nineteen testmore » series and a total of 178 tests were performed. Testing addressed the effects of: Heat flux; pressure; helium gas; power tilt; ribs; asymmetric heat flux. This document consists solely of the plato file index from 11/87 to 11/90.« less
Publication Date:
OSTI Identifier:
10148047
Report Number(s):
WSRC-TR--92-534-Bk.5
ON: DE94011254; TRN: 94:011987
DOE Contract Number:
AC09-89SR18035
Resource Type:
Technical Report
Resource Relation:
Other Information: PBD: Nov 1992
Research Org:
Westinghouse Savannah River Co., Aiken, SC (United States)
Sponsoring Org:
USDOE, Washington, DC (United States)
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; SPECIAL PRODUCTION REACTORS; REACTOR COOLING SYSTEMS; TESTING; COMPILED DATA; LOSS OF COOLANT; HYDRAULICS; HEAT TRANSFER; NUCLEATE BOILING; REACTOR SAFETY 220900; 220600; REACTOR SAFETY; RESEARCH, TEST, TRAINING, PRODUCTION, IRRADIATION, MATERIALS TESTING REACTORS