Power Burst Facility (PBF) severe fuel damage test 1-4 test results report

Technical Report · Sat Apr 01 04:00:00 EST 1989

DOI:https://doi.org/10.2172/5983389· OSTI ID:5983389

Petti, D A; Martinson, Z R; Hobbins, R R; Allison, C M; Carlson, E R; Hagrman, D L; Cheng, T C; Hartwell, J K; Vinjamuri, K; Seifken, L J

A comprehensive evaluation of the Severe Fuel Damage (SFD) Test 1-4 performed in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory is presented. Test SFD 1-4 was the fourth and final 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 and control rod behavior, aerosol and hydrogen generation, and fission product release and transport during a high-temperature, severe fuel damage transient. A test bundle, comprised of 26 previously irradiated (36,000 MWd/MtU) pressurized water-reactor-type fuel rods, 2 fresh instrumented fuel rods, and 4 silver-indium-cadmium control rods, was surrounded by an insulating shroud and contained in a pressurized in-pile tube. The experiment consisted of a 1.3-h transient at a coolant pressure of 6.95 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 and control rod absorber alloy melting, fuel liquefaction, material relocation, and the release of hydrogen, aerosols, and fission products. The transient was terminated over a 2100-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 on-line instrumentation, analysis of fission product and aerosol data, postirradiation examination of the fuel bundle, and calculations using the SCDAP/RELAP5 computer code, are presented. 40 refs., 160 figs., 31 tabs.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (USA). Div. of Systems Research; EG and G Idaho, Inc., Idaho Falls, ID (USA)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 5983389
• Report Number(s): NUREG/CR-5163; EGG-2542; ON: TI89012584
• Country of Publication: United States
• Language: English

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22 GENERAL STUDIES OF NUCLEAR REACTORS
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ACCIDENTS
AEROSOLS
ALLOYS
BWR TYPE REACTORS
CHEMICAL REACTIONS
COLLOIDS
COMPUTER CODES
CONTROL ELEMENTS
DAMAGE
DATA
DISPERSIONS
ELEMENTS
ENERGY TRANSFER
EXPERIMENTAL DATA
FISSION PRODUCT RELEASE
FLUID MECHANICS
FUEL ELEMENTS
HEAT TRANSFER
HYDRAULICS
HYDROGEN
INFORMATION
LOSS OF FLOW
MATERIALS
MECHANICS
NONMETALS
NUMERICAL DATA
OXIDATION
PBF REACTOR
POST-IRRADIATION EXAMINATION
PULSED REACTORS
PWR TYPE REACTORS
RADIATION TRANSPORT
RADIOACTIVE EFFLUENTS
RADIOACTIVE MATERIALS
RADIOACTIVE WASTES
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SAFETY
REACTORS
S CODES
SAFETY
SOLS
TANK TYPE REACTORS
TESTING
WASTES
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS