Validation of the HOTCHAN code for analyzing the EBR-II core following an unprotected loss of flow

Conference · Fri Jan 01 04:00:00 EST 1988 · Trans. Am. Nucl. Soc.; (United States)

OSTI ID:5852258

Mohr, D; Chang, L K; Planchon, H P

A series of reactor experiments involving unprotected (no scram) loss-of-primary flow (LOF) down to natural convection was successfully conducted in February and April of 1986 on the Experimental Breeder Reactor II (EBR-II). The predicted and measured behavior of a special instrumented assembly, the XX09-fueled INSAT, was compared for the most severe test (SHRT 45) to demonstrate the validation of the thermal-hydraulic code HOTCHAN. The particular test of interest in this paper was initiated at full power by tripping the primary and secondary pumps. These tests were part of the shutdown heat removal tests (SHRT) being conducted in EBR-II. The reactor and balance of plant are extensively instrumented, and measurements were recorded by a data acquisition system. The reactor and plant response confirm predictions that the driver fuel cladding can survive temperatures above the eutectic threshold for the transient following a station blackout without scramming the reactor. In addition, the in-core data provide a firm basis for validation of the Argonne/EBR-II developed HOTCHAN code for analyzing the thermal-hydraulic behavior of specific fuel subassemblies. In this paper the analytical model for HOTCHAN is described as well as it relationship to the NATDEMO code. The predicted behavior of the hottest driver subassembly is also discussed and compared with the INSAT XX09 results.

• Research Organization: Argonne National Lab., IL (USA)
• OSTI ID: 5852258
• Report Number(s): CONF-881011-
• Conference Information: Journal Name: Trans. Am. Nucl. Soc.; (United States) Journal Volume: 57
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210500 -- Power Reactors
Breeding
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ACTINIDES
ALGORITHMS
ALKALI METALS
BREEDER REACTORS
COMPUTER CODES
COMPUTERIZED SIMULATION
CONVECTION
COOLING SYSTEMS
DECAY
DEPLETED URANIUM
EBR-2 REACTOR
ELEMENTS
ENERGY SYSTEMS
ENERGY TRANSFER
EPITHERMAL REACTORS
EXPERIMENTAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FLUID MECHANICS
FUEL ASSEMBLIES
H CODES
HEAT TRANSFER
HYDRAULICS
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
LOSS OF FLOW
MASS TRANSFER
MATHEMATICAL LOGIC
MECHANICS
METALS
N CODES
NATURAL CONVECTION
OPERATION
OUTAGES
PERFORMANCE TESTING
PLUTONIUM
POWER DENSITY
POWER REACTORS
PRESSURE DROP
PRIMARY COOLANT CIRCUITS
PUMPS
REACTIVITY
REACTOR ACCIDENTS
REACTOR CHANNELS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR CORES
REACTOR OPERATION
REACTOR SAFETY EXPERIMENTS
REACTOR SHUTDOWN
REACTORS
RESEARCH AND TEST REACTORS
RHR SYSTEMS
SCRAM
SHUTDOWNS
SIMULATION
SODIUM
SODIUM COOLED REACTORS
TESTING
TRANSIENTS
TRANSURANIUM ELEMENTS
URANIUM
VALIDATION