Validation of ATR SINDA using SPERT-3 reactivity-initiated transient test data

Conference · Tue Jan 01 00:00:00 EST 1991 · Transactions of the American Nuclear Society; (United States)

OSTI ID:5776548

Oh, C H; Ambrosek, R G; Wadkins, R P ^[1]

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1. Idaho National Engineering Lab., Idaho Falls (United States)

Thermal-hydraulic analyses of Special Power Excursion Reactor Test-3 (SPERT-3) reactivity-initiated accident (RIA) test data are presented to validate ATR SINDA, a three-dimensional thermal-hydraulic computer code used to conduct transient analysis of Advanced Test Reactor (ATR) fuel plates. Because SPERT-3-C core test results are the only existing data of reactivity insertions for plate-type reactors, validating ATR SINDA is important for safety analysis of postulated reactivity initiated accidents (RIAs) for the plate-type ATR operated at the Idaho National Engineering Laboratory (INEL). The SPERT-3 facility was designed to study the kinetic behavior and safety of pressurized, light water-moderated reactors.

• OSTI ID: 5776548
• Report Number(s): CONF-910603-; CODEN: TANSA
• Journal Information: Transactions of the American Nuclear Society; (United States), Vol. 63; Conference: Annual meeting of the American Nuclear Society (ANS), Orlando, FL (United States), 2-6 Jun 1991; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
42 ENGINEERING
ATR REACTOR
EXCURSIONS
REACTOR SAFETY
S CODES
COMPARATIVE EVALUATIONS
COMPUTER CODES
COMPUTERIZED SIMULATION
CONTROL ELEMENTS
EVALUATED DATA
EXPERIMENTAL DATA
FUEL ASSEMBLIES
FUEL CANS
FUEL PLATES
HEAT TRANSFER
HYDRAULICS
IDAHO NATIONAL ENGINEERING LABORATORY
PWR TYPE REACTORS
REACTIVITY INSERTIONS
REACTOR ACCIDENTS
REACTOR KINETICS
SENSITIVITY ANALYSIS
TEST FACILITIES
THREE-DIMENSIONAL CALCULATIONS
TRANSIENTS
VALIDATION
ACCIDENTS
DATA
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
EVALUATION
FLUID MECHANICS
FUEL ELEMENTS
INFORMATION
IRRADIATION REACTORS
KINETICS
MATERIALS TESTING REACTORS
MECHANICS
NATIONAL ORGANIZATIONS
NUMERICAL DATA
POWER REACTORS
REACTOR COMPONENTS
REACTORS
RESEARCH AND TEST REACTORS
SAFETY
SIMULATION
TANK TYPE REACTORS
TEST REACTORS
TESTING
THERMAL REACTORS
US DOE
US ERDA
US ORGANIZATIONS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
220600 - Nuclear Reactor Technology- Research
Test & Experimental Reactors
420400 - Engineering- Heat Transfer & Fluid Flow