SASSYS validation with the EBR-II shutdown heat removal tests

Conference · Sun Jan 01 00:00:00 EST 1989

OSTI ID:5496892

Herzog, J P ^[1]

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1. Argonne National Lab., IL (USA)

SASSYS is a coupled neutronic and thermal hydraulic code developed for the analysis of transients in liquid metal cooled reactors (LMRs). The code is especially suited for evaluating of normal reactor transients -- protected (design basis) and unprotected (anticipated transient without scram) transients. Because SASSYS is heavily used in support of the IFR concept and of innovative LMR designs, such as PRISM, a strong validation base for the code must exist. Part of the validation process for SASSYS is analysis of experiments performed on operating reactors, such as the metal fueled Experimental Breeder Reactor -- II (EBR-II). During the course of a series of historic whole-plant experiments, EBR-II illustrated key safety features of metal fueled LMRs. These experiments, the Shutdown Heat Removal Tests (SHRT), culminated in unprotected loss of flow and loss of heat sink transients from full power and flow. Analysis of these and earlier SHRT experiments constitutes a vital part of SASSYS validation, because it facilitates scrutiny of specific SASSYS models and of integrated code capability. 12 refs., 11 figs.

• Research Organization: Argonne National Lab. (ANL), Argonne, IL (United States)
• Sponsoring Organization: DOE/NE
• DOE Contract Number: W-31109-ENG-38
• OSTI ID: 5496892
• Report Number(s): CONF-891140-3; ON: DE90001924; TRN: 89-030984
• Resource Relation: Conference: Session on thermal-hydraulic aspects of passive safety and new generalization reactors, San Francisco, CA (USA), 27-30 Nov 1989
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE
EBR-2 REACTOR
LOSS OF COOLANT
TESTING
BLACKOUTS
COMPARATIVE EVALUATIONS
DATA ANALYSIS
DESIGN
FEEDBACK
FLUID FLOW
GEOMETRY
HEAT TRANSFER
HYDRAULICS
POWER GENERATION
PUMPS
REACTOR COOLING SYSTEMS
REACTOR CORES
REACTOR LICENSING
REACTOR SAFETY
REACTOR SHUTDOWN
S CODES
SCRAM
TEMPERATURE EFFECTS
TRANSIENTS
ACCIDENTS
BREEDER REACTORS
COMPUTER CODES
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
EPITHERMAL REACTORS
EXPERIMENTAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FLUID MECHANICS
LICENSING
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
MATHEMATICS
MECHANICS
POWER REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
RESEARCH AND TEST REACTORS
SAFETY
SHUTDOWN
SODIUM COOLED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
220600 - Nuclear Reactor Technology- Research
Test & Experimental Reactors
210500 - Power Reactors
Breeding
990220 - Computers
Computerized Models
& Computer Programs- (1987-1989)