Fuel performance during severe accidents. [PWR]
Conference
·
OSTI ID:6977877
As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. This program is underway in the Power Burst Facility at the Idaho National Engineering Laboratory. In preparation for the first test, predictions have been performed using the TRAC-BD1 computer. This paper presents the calculated results showing a slow heatup to 2400 K over 5 hours, and the analysis includes accelerated oxidation of the zirconium cladding at temperatures above 1850 K.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 6977877
- Report Number(s):
- EGG-M-04082; CONF-820802-41; ON: DE83000545
- Country of Publication:
- United States
- Language:
- English
Similar Records
Severe fuel-damage scoping test performance. [PWR]
Severe fuel damage scoping test performance
Power Burst Facility severe-fuel-damage test program. [PWR]
Conference
·
Fri Dec 31 23:00:00 EST 1982
·
OSTI ID:5624709
Severe fuel damage scoping test performance
Journal Article
·
Fri Dec 31 23:00:00 EST 1982
· AIChE Symp. Ser.; (United States)
·
OSTI ID:5375866
Power Burst Facility severe-fuel-damage test program. [PWR]
Conference
·
Thu Dec 31 23:00:00 EST 1981
·
OSTI ID:6797093
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
ALLOYS
CHEMICAL REACTIONS
COMPUTER CALCULATIONS
DAMAGE
FUEL ASSEMBLIES
FUEL CANS
FUEL ELEMENTS
FUEL RODS
OXIDATION
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SAFETY
REACTORS
SAFETY
STRESSES
TEMPERATURE GRADIENTS
THERMAL STRESSES
TIN ALLOYS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ALLOYS
CHEMICAL REACTIONS
COMPUTER CALCULATIONS
DAMAGE
FUEL ASSEMBLIES
FUEL CANS
FUEL ELEMENTS
FUEL RODS
OXIDATION
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SAFETY
REACTORS
SAFETY
STRESSES
TEMPERATURE GRADIENTS
THERMAL STRESSES
TIN ALLOYS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS