Experimental investigation of thermal limits in parallel plate configuration for the Advanced Neutron Source Reactor

Conference · Mon Mar 01 04:00:00 EST 1993

OSTI ID:10133138

Siman-Tov, M; Felde, D K; Kaminaga, M; Yoder, G L

The Advanced Neutron Source Reactor (ANSR) is currently being designed to become the world`s highest-flux, steady-state, thermal neutron source for scientific experiments. Highly subcooled, heavy-water coolant flows vertically upward at a very high velocity of 25 m/s through parallel aluminum fuel-plates. The core has average and peak heat fluxes of 5.9 and 12 MW/m{sup 2}, respectively. In this configuration, both flow excursion (FE) and true critical heat flux (CHF), represent potential thermal limitations. The availability of experimental data for both FE and true CHF at the conditions applicable to the ANSR is very limited. A Thermal Hydraulic Test Loop (THTL) facility was designed and built to simulate a full-length coolant subchannel of the core, allowing experimental determination of both thermal limits under the expected ANSR T/H conditions. A series of FE tests with water flowing vertically upward was completed over a nominal heat flux range of 6 to 14 MW/m{sup 2} and a corresponding velocity range of 8 to 21 m/s. Both the exit pressure (1.7 MPa) and inlet temperature (45{degrees}C) were maintained constant for these tests, while the loop was operated in a ``stiff``(constant flow) mode. Limited experiments were also conducted at 12 MW/m{sup 2} using a ``soft`` mode (near constant pressure-drop) for actual FE burnout tests and using a ``stiff` mode for true CHF tests, to compare with the original FE experiments.

• Research Organization: Oak Ridge National Lab., TN (United States)
• Sponsoring Organization: USDOE, Washington, DC (United States)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 10133138
• Report Number(s): CONF-930830--6; ON: DE93008274
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220300
220600
FLUID FLOW
FUEL CHANNELS
FUEL ELEMENTS
FUEL PLATES
HEAT TRANSFER
HYDRAULICS
LIMITING VALUES
REACTOR COOLING SYSTEMS
RESEARCH REACTORS
RESEARCH
TEST
TRAINING
PRODUCTION
IRRADIATION
MATERIALS TESTING REACTORS
TEST FACILITIES
TESTING
THERMAL ANALYSIS