RELAP5 assessment: semiscale Mod-3 small-break tests. [PWR]
Technical Report
·
OSTI ID:5620951
The RELAP5 independent assessment project at Sandia National Laboratories is part of an overall effort funded by the NRC to determine the ability of various systems codes to predict the detailed thermal-hydraulic response of LWR's during accident and off-normal conditions. The RELAP5/MOD1 code is being assessed at SNLA against test data from various integral and separate effects test facilities. As part of this assessment matrix, a series of small break experiments performed at the Semiscale Mod-3 facility have been analyzed. The results show that RELAP5/MOD1 will predict reasonably well many aspects of the four small break transients analyzed. Many problems were encountered in matching the initial primary and secondary steady state conditions, which in turn affected the transient calculations. Although the calculations predicted the relative severity of the experiments in terms of mass inventories and distribution, quantitatively they generally predicted a more severe transient than was observed experimentally.
- Research Organization:
- Sandia National Labs., Albuquerque, NM (USA)
- DOE Contract Number:
- AC04-76DP00789
- OSTI ID:
- 5620951
- Report Number(s):
- NUREG/CR-3277; SAND-83-1038; ON: DE84000175
- Country of Publication:
- United States
- Language:
- English
Similar Records
RELAP5 assessment: LOFT small-break L3-6/L8-1. [PWR]
RELAP5 assessment: LOFT intermediate breaks L5-1 and L8-2. [PWR]
RELAP5 assessment: semiscale natural circulation tests S-NC-3, S-NC-4, and S-NC-8
Technical Report
·
Mon Feb 28 23:00:00 EST 1983
·
OSTI ID:6016012
RELAP5 assessment: LOFT intermediate breaks L5-1 and L8-2. [PWR]
Technical Report
·
Mon Aug 01 00:00:00 EDT 1983
·
OSTI ID:5539379
RELAP5 assessment: semiscale natural circulation tests S-NC-3, S-NC-4, and S-NC-8
Technical Report
·
Tue May 01 00:00:00 EDT 1984
·
OSTI ID:6894374
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
COMPUTER CALCULATIONS
DATA
ENERGY TRANSFER
EXPERIMENTAL DATA
FLOW RATE
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INFORMATION
LOSS OF COOLANT
MECHANICS
NUMERICAL DATA
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
TEMPERATURE GRADIENTS
TEST FACILITIES
THEORETICAL DATA
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 CALCULATIONS
DATA
ENERGY TRANSFER
EXPERIMENTAL DATA
FLOW RATE
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INFORMATION
LOSS OF COOLANT
MECHANICS
NUMERICAL DATA
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
TEMPERATURE GRADIENTS
TEST FACILITIES
THEORETICAL DATA
WATER COOLED REACTORS
WATER MODERATED REACTORS