Thermal/Hydraulic Analysis Research Program. Quarterly report, January-March 1984. Volume 1 of 4. [PWR]
Technical Report
·
OSTI ID:6647055
The TRAC-PF1/MOD1 independent assessment program at Sandia National Laboratories (SNLA) is part of a multi-faceted effort sponsored by the Nuclear Regulatory Commission (NRC) to determine the ability of various systems codes to predict the detailed thermal/hydraulic response of LWRs during accident and off-normal conditions. The first quarter of FY84 marked the beginning of the TRAC-PF1/MOD1 independent assessment project at SNLA. The code was obtained from Loss Alamos National Laboratory (LANL) in October, and brought up on both our Cyber-76 and Cray-1S computers. The assessment matrix was formalized, several TRAC nodalizations for the various facilities required were developed, and limited calculations were begun, all described in the last quarterly. During this quarter, more nodalizations were developed and calculations begun, and the first PF1/MOD1 assessment analysis was completed.
- Research Organization:
- Sandia National Labs., Albuquerque, NM (USA)
- DOE Contract Number:
- AC04-76DP00789
- OSTI ID:
- 6647055
- Report Number(s):
- NUREG/CR-3820; SAND-84-1025/1; ON: TI84015548
- Country of Publication:
- United States
- Language:
- English
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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
ACCIDENTS
BOILERS
COMPUTER CALCULATIONS
DATA
ENERGY TRANSFER
EXPERIMENTAL DATA
FLOW RATE
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INFORMATION
LOSS OF FLOW
MECHANICS
NUMERICAL DATA
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
STEAM GENERATORS
TEMPERATURE GRADIENTS
THEORETICAL DATA
VAPOR GENERATORS
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
BOILERS
COMPUTER CALCULATIONS
DATA
ENERGY TRANSFER
EXPERIMENTAL DATA
FLOW RATE
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INFORMATION
LOSS OF FLOW
MECHANICS
NUMERICAL DATA
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
STEAM GENERATORS
TEMPERATURE GRADIENTS
THEORETICAL DATA
VAPOR GENERATORS
WATER COOLED REACTORS
WATER MODERATED REACTORS