PBF (Power Burst Facility) severe fuel damage test 1--3 test results report
- EG and G Idaho, Inc., Idaho Falls, ID (USA)
A comprehensive evaluation of the Severe Fuel Damage (SFD) Test 1--3 performed in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory is presented. Test SFD 1--3 was the third test in an internationally sponsored light water reactor severe accident research program, initiated by the US Nuclear Regulatory Commission. The overall technical objective of the test was to contribute to the understanding of fuel rod behavior, hydrogen generation, and fission product release and transport during a high-temperature, severe fuel damage transient. A test bundle, comprised of 26 previously irradiated (38,000 MWd/tU) pressurized water reactor-type fuel rods, 2 fresh instrumented fuel rods, and 4 empty zircaloy guide tubes, was surrounded by an insulating shroud and contained in a pressurized in-pile tube. The experiment consisted of a 1-h transient at a nominal coolant pressure of 6.85 MPa in which the inlet coolant flow to the bundle was reduced to 0.6 g/s while the bundle fission power was gradually increased until dryout, heatup, cladding rupture, and oxidation occurred. With sustained fission power and heat from oxidation, temperatures continued to rise rapidly, resulting in zircaloy melting, fuel liquefaction, material relocation, and the release of hydrogen, aerosols, and fission products. The transient was terminated over a 1340-s time span by slowly reducing the reactor power and cooling the damaged bundle with argon gas. A description and evaluation of the major phenomena, based upon the response of online instrumentation, analysis of fission product data, postirradiation examination of the fuel bundle, and calculations using the SCDAP/RELAP5 computer code, are presented. 34 refs., 241 figs., 51 tabs.
- Research Organization:
- Nuclear Regulatory Commission, Washington, DC (USA). Div. of Systems Research; EG and G Idaho, Inc., Idaho Falls, ID (USA)
- Sponsoring Organization:
- USNRC
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 5321163
- Report Number(s):
- NUREG/CR-5354; EGG-2565; ON: TI90004212; TRN: 90-004079
- Resource Relation:
- Other Information: This report contains 6 microfiche supplements
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE
PBF REACTOR
FUEL ELEMENTS
COMPARATIVE EVALUATIONS
COMPUTERIZED SIMULATION
CONTROL ROD DRIVES
EXPERIMENTAL DATA
FISSION PRODUCT RELEASE
GAMMA SPECTROSCOPY
HEAT TRANSFER
HYDRAULICS
HYDROGEN
IRRADIATION
MELTDOWN
PERFORMANCE TESTING
PRESSURE EFFECTS
RADIATION HAZARDS
RADIATION MONITORING
REACTOR ACCIDENTS
REACTOR CORES
REACTOR INSTRUMENTATION
REACTOR SAFETY
S CODES
TEMPERATURE EFFECTS
TRANSIENTS
ZIRCALOY
ACCIDENTS
ALLOYS
COMPUTER CODES
DATA
ELEMENTS
ENERGY TRANSFER
FLUID MECHANICS
HAZARDS
HEALTH HAZARDS
INFORMATION
MECHANICS
MONITORING
NONMETALS
NUMERICAL DATA
PULSED REACTORS
REACTOR COMPONENTS
REACTORS
SAFETY
SIMULATION
SPECTROSCOPY
TANK TYPE REACTORS
TESTING
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
220900* - Nuclear Reactor Technology- Reactor Safety
220600 - Nuclear Reactor Technology- Research
Test & Experimental Reactors
990200 - Mathematics & Computers