TRAC-PF1/MOD1 assessment at Los Alamos
Conference
·
OSTI ID:6409388
The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in pressurized water reactors (PWRs). Over the past several years, four distinct versions of the code have been released; each new version introduced improvements to the existing models and numerics and added new models to extend the applications of the code. The first goal of the code was to analyze large-break loss-of-coolant accidents (LOCAs), and the TRAC-P1A and TRAC-PD2 codes primarily addressed the large-break LOCA. (The TRAC-PD2/MOD1 code is essentially the same as the TRAC-PD2 code but it also includes a released set of error corrections.) The TRAC-PF1 code contained major changes to the models and trips and to the numerical methods. These modifications enhanced the computational speed of the code and improved the application to small-break LOCAs. The TRAC-PF1/MOD1 code, the latest released version, added improved steam-generator modeling, a turbine component, and a control system together with modified constitutive relations to model the balance of plant on the secondary side and to extend the applications to non-LOCA transients. The TRAC-PF1/MOD1 code also contains reasonably general reactor-kinetics modeling to facilitate the simulation of transients with delayed scram or without scram. 13 references, 24 figures.
- Research Organization:
- Los Alamos National Lab., NM (USA)
- DOE Contract Number:
- W-7405-ENG-36
- OSTI ID:
- 6409388
- Report Number(s):
- LA-UR-84-3316; CONF-8410142-23; ON: DE85002380
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
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210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
COMPUTER CODES
COMPUTERIZED SIMULATION
EVALUATION
LOSS OF COOLANT
MATHEMATICAL MODELS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTORS
SIMULATION
T CODES
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
COMPUTER CODES
COMPUTERIZED SIMULATION
EVALUATION
LOSS OF COOLANT
MATHEMATICAL MODELS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTORS
SIMULATION
T CODES
WATER COOLED REACTORS
WATER MODERATED REACTORS