RELAP5/MOD3 Code Manual

Technical Report · Wed Jan 01 04:00:00 EST 1992

OSTI ID:5559769

Fletcher, C D; Schultz, R R ^[1]

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1. EG and G Idaho, Inc., Idaho Falls, ID (United States)

The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system and the core during the severe accident transient, as well as large and small break loss-of-coolant accidents and operational transients, such as anticipated transients without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. RELAP5/MOD3 code documentation is divided into five volumes: Volume 1 describes modeling theory and associated numerical schemes, Volume 2 contains detailed instructions for code application and input data preparation, Volume 3 provides the results of developmental assessment cases that demonstrate and verify the models used in the code, and Volume 4 presents a detailed discussion of RELAP5 models and correlations. Volumes 1--4 are in varying stages of development. This document, Volume 5, contains guidelines that have evolved over the past several years through use of the RELAP5 code.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research; EG and G Idaho, Inc., Idaho Falls, ID (United States)
• Sponsoring Organization: NRC; Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 5559769
• Report Number(s): NUREG/CR-5535-Vol.5; EGG--2596-Vol.5; ON: TI92010298
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
99 GENERAL AND MISCELLANEOUS
990200 -- Mathematics & Computers
ACCIDENTS
ATWS
BLACKOUTS
COMPUTER CODES
COMPUTERIZED SIMULATION
COOLING SYSTEMS
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
LOSS OF FLOW
MECHANICS
POWER REACTORS
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR CORES
REACTORS
SIMULATION
STANDARDS
THERMAL REACTORS
TRANSIENTS
WATER COOLED REACTORS
WATER MODERATED REACTORS