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

Abstract

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 ofmore » 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.« less

Authors:
; ; ; ; ; ; ; ; ;
Publication Date:
Research Org.:
Nuclear Regulatory Commission, Washington, DC (USA). Div. of Systems Research; EG and G Idaho, Inc., Idaho Falls, ID (USA)
OSTI Identifier:
5983389
Report Number(s):
NUREG/CR-5163; EGG-2542
ON: TI89012584
DOE Contract Number:  
AC07-76ID01570
Resource Type:
Technical Report
Resource Relation:
Other Information: Includes 9 sheets of 24X reduction microfiche
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; PBF REACTOR; FUEL ELEMENTS; AEROSOLS; BWR TYPE REACTORS; CONTROL ELEMENTS; DAMAGE; EXPERIMENTAL DATA; FISSION PRODUCT RELEASE; HEAT TRANSFER; HYDRAULICS; HYDROGEN; LOSS OF FLOW; OXIDATION; POST-IRRADIATION EXAMINATION; PWR TYPE REACTORS; RADIATION TRANSPORT; RADIOACTIVE EFFLUENTS; REACTOR SAFETY; S CODES; TESTING; ZIRCALOY; ACCIDENTS; ALLOYS; CHEMICAL REACTIONS; COLLOIDS; COMPUTER CODES; DATA; DISPERSIONS; ELEMENTS; ENERGY TRANSFER; FLUID MECHANICS; INFORMATION; MATERIALS; MECHANICS; NONMETALS; NUMERICAL DATA; PULSED REACTORS; RADIOACTIVE MATERIALS; RADIOACTIVE WASTES; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTORS; SAFETY; SOLS; TANK TYPE REACTORS; WASTES; WATER COOLED REACTORS; WATER MODERATED REACTORS; ZIRCONIUM ALLOYS; ZIRCONIUM BASE ALLOYS; 220900* - Nuclear Reactor Technology- Reactor Safety; 220502 - Nuclear Reactor Technology- Environmental Aspects- Radioactive Effluents; 210200 - Power Reactors, Nonbreeding, Light-Water Moderated, Nonboiling Water Cooled; 210100 - Power Reactors, Nonbreeding, Light-Water Moderated, Boiling Water Cooled; 220600 - Nuclear Reactor Technology- Research, Test & Experimental Reactors

Citation Formats

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., and Seifken, L.J. Power Burst Facility (PBF) severe fuel damage test 1-4 test results report. United States: N. p., 1989. Web. doi:10.2172/5983389.
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. Power Burst Facility (PBF) severe fuel damage test 1-4 test results report. United States. doi:10.2172/5983389.
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., and Seifken, L.J. Sat . "Power Burst Facility (PBF) severe fuel damage test 1-4 test results report". United States. doi:10.2172/5983389. https://www.osti.gov/servlets/purl/5983389.
@article{osti_5983389,
title = {Power Burst Facility (PBF) severe fuel damage test 1-4 test results report},
author = {Petti, D.A. and Martinson, Z.R. and Hobbins, R.R. and Allison, C.M. and Carlson, E.R. and Hagrman, D.L. and Cheng, T.C. and Hartwell, J.K. and Vinjamuri, K. and Seifken, L.J.},
abstractNote = {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.},
doi = {10.2172/5983389},
journal = {},
number = ,
volume = ,
place = {United States},
year = {1989},
month = {4}
}