The Advanced Neutron Source Thermal Hydraulic Test Loop

Conference · Fri May 01 04:00:00 EDT 1992

OSTI ID:10137700

Felde, D K; Montgomery, B H; Yoder, G L ^[1]; Skrzycke, D P ^[2]

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1. Oak Ridge National Lab., TN (United States)
2. Tennessee Univ., Knoxville, TN (United States). Dept. of Mechanical Engineering

The Advanced Neutron Source (ANS) Thermal Hydraulic Test Loop has been designed to provide known thermal-hydraulic conditions to a test section which simulates a full-length aluminum coolant subchannel of the ANS reactor core. Tests in this facility will provide both single- and two-phase thermal-hydraulic information. The specific phenomena that are to be examined are (1) single-phase heat transfer coefficients and friction factors, (2) the point of incipient boiling, (3) nucleate boiling heat transfer coefficients, (4) two-phase pressure drop characteristics in the nucleate boiling regime, (5) flow instability limits, and (6) critical heat flux limits. A Moyno ``near`` positive displacement type primary group with a 4 MPa differential head is capable of providing up to 35 m/s flow through the narrow channel test section. Two 12-volt dc power supplies, operated in parallel, with 40 kA total current capacity provide direct resistance heating to the test channel. A primary heat exchanger is designed to remove up to 410 kw. Data acquisition for the loop is through two PC-based systems, one of which is a fast data acquisition system capable of measuring up to 100,000 samples/s. As an aid in test design and analysis, an analytical model has been developed to predict the thermal hydraulic response of the test loop under transient conditions.

• Research Organization: Oak Ridge National Lab., TN (United States)
• Sponsoring Organization: USDOE, Washington, DC (United States)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 10137700
• Report Number(s): CONF-920538--11; ON: DE92011105
• Country of Publication: United States
• Language: English

Related Subjects

07 ISOTOPE AND RADIATION SOURCES
070201
22 GENERAL STUDIES OF NUCLEAR REACTORS
220600
ALUMINIUM
DATA ACQUISITION SYSTEMS
DESIGN, FABRICATION, AND OPERATION
HEAT TRANSFER
HYDRAULICS
NEUTRON SOURCE FACILITIES
NEUTRON SOURCES
NUCLEATE BOILING
REACTOR COOLING SYSTEMS
RESEARCH REACTORS
RESEARCH
TEST
TRAINING
PRODUCTION
IRRADIATION
MATERIALS TESTING REACTORS
TESTING