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Title: METAL-WATER REACTIONS: III. FUEL ELEMENT STRESSES DURING A NUCLEAR ACCIDENT

Abstract

The thermal and pressure stresses in metallic plate and rod type nuclear fuel elements during a nuclear accident severe enough tc produce a metal-water reaction have been reviewed. The initial reactor pericds required to cause these severe accidents are in the range of 4 to 12 milliseconds. The result of the stress review presented in this report indicates that the metallic plate type fuel will swell from internal fission gas pressure, the rate of gas generation, and the magnitude and exposure to elevated central temperatures. Pulsed irradiation type experiments on sample fuel elements are desirable to determine the extent and consequences of this swelling including determination of the mechanism of fuel element failure and if dispersion occurs, measurement of the size of the particles. The stress review for the clad oxide rod type fuel, indicates that the oxide core will expand more than the Zircaloy cladding during severe reactor accidents producing cladding strains in the range of one to two percent. Experimentally, 7 to 10 percent strain is required to rupture irradiated Zircaloy tubing indicating that the cladding should not fail during the reactor transients considered. However, localized areas of the cladding may experience damage during fabrication, loss of ductilitymore » or strain concentrations sufficient to crack the cladding. There is no evidence to indicate that the cladding will be pulverized into small enough pieces to accelerate a metalwater reaction nor to modify the analytical models used to evaluate the core conditions during the accidents considered. Pulsed irradiation type experiments on fuel specimens simulating the conditions reviewed would be desirable to evaluate the transient material and mechanical properties of the fuel elements. (auth)« less

Authors:
Publication Date:
Research Org.:
General Electric Co. Atomic Power Equipment Dept., San Jose, Calif.
Sponsoring Org.:
US Atomic Energy Commission (AEC)
OSTI Identifier:
4210187
Report Number(s):
GEAP-3191
NSA Number:
NSA-14-002140
DOE Contract Number:  
AT(04-3)-189
Resource Type:
Technical Report
Resource Relation:
Other Information: Orig. Receipt Date: 31-DEC-60
Country of Publication:
United States
Language:
English
Subject:
REACTOR TECHNOLOGY; ACCIDENTS; DISPERSIONS; EXCURSIONS; FAILURES; FISSION PRODUCTS; FUEL CANS; FUEL ELEMENTS; GASES; HIGH TEMPERATURE; IRRADIATION; METALS; MOCKUP; OXIDES; PLATES; PRESSURE; REACTIVITY; REACTOR CORE; REACTOR SAFETY; REACTORS; RODS; STRESSES; TEMPERATURE; TESTING; WATER; ZIRCALOY

Citation Formats

Horst, K. M. METAL-WATER REACTIONS: III. FUEL ELEMENT STRESSES DURING A NUCLEAR ACCIDENT. United States: N. p., 1959. Web. doi:10.2172/4210187.
Horst, K. M. METAL-WATER REACTIONS: III. FUEL ELEMENT STRESSES DURING A NUCLEAR ACCIDENT. United States. doi:10.2172/4210187.
Horst, K. M. Fri . "METAL-WATER REACTIONS: III. FUEL ELEMENT STRESSES DURING A NUCLEAR ACCIDENT". United States. doi:10.2172/4210187. https://www.osti.gov/servlets/purl/4210187.
@article{osti_4210187,
title = {METAL-WATER REACTIONS: III. FUEL ELEMENT STRESSES DURING A NUCLEAR ACCIDENT},
author = {Horst, K. M.},
abstractNote = {The thermal and pressure stresses in metallic plate and rod type nuclear fuel elements during a nuclear accident severe enough tc produce a metal-water reaction have been reviewed. The initial reactor pericds required to cause these severe accidents are in the range of 4 to 12 milliseconds. The result of the stress review presented in this report indicates that the metallic plate type fuel will swell from internal fission gas pressure, the rate of gas generation, and the magnitude and exposure to elevated central temperatures. Pulsed irradiation type experiments on sample fuel elements are desirable to determine the extent and consequences of this swelling including determination of the mechanism of fuel element failure and if dispersion occurs, measurement of the size of the particles. The stress review for the clad oxide rod type fuel, indicates that the oxide core will expand more than the Zircaloy cladding during severe reactor accidents producing cladding strains in the range of one to two percent. Experimentally, 7 to 10 percent strain is required to rupture irradiated Zircaloy tubing indicating that the cladding should not fail during the reactor transients considered. However, localized areas of the cladding may experience damage during fabrication, loss of ductility or strain concentrations sufficient to crack the cladding. There is no evidence to indicate that the cladding will be pulverized into small enough pieces to accelerate a metalwater reaction nor to modify the analytical models used to evaluate the core conditions during the accidents considered. Pulsed irradiation type experiments on fuel specimens simulating the conditions reviewed would be desirable to evaluate the transient material and mechanical properties of the fuel elements. (auth)},
doi = {10.2172/4210187},
journal = {},
number = ,
volume = ,
place = {United States},
year = {1959},
month = {7}
}