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Title: A STUDY OF URANIUM CARBIDE AND CLADDING MATERIALS FOR HIGH-TEMPERATURE SODIUM-COOLED REACTORS

Abstract

Irradiation experiments were conducted to determine the irradiation performance characteristics of uranium monocarbide, exposed to high-temperature ( approximates 2000 deg F), high heat flux ( approximates -10/sup 6/ Btu/hr-ft/sup 2/), high-burnup ( approximates 25,000 Mwd/ MTU) operation. Hyperstoichiometric uranium carbide fuels have sustained maximum conditions of burnup and average central temperature of 24,000 and 10,000 Mwd/ MTU at 1550 and 2150 deg F, respectively. Hypostoichiometric uranium carbide fuels were exposed to conditions of 1800 deg F average central temperature for a burnup of approximates 7000 Mwd/ MTU. Fission gas release, in general, was low. Values for fuel dimensional increases were scattered, due to thermal or mechanical effects over the fuel column lengths, and could not be correlated with burnup or average fuel central temperatures. The fuel density decrease profile of the fuel length was similar to the burnup profile. Hyperstoichiometric UC/Na (or NaK) Type 304 stainless steel systems show excessive cladding carburization at fuel surface temperatures as low as 1000 deg F for an acceptable reactor residence time. Hypostoichiometric UC/Na Type 304 stainless steel systems are compatible at 1200 and 1600 deg F under isothermal conditions for at least 6400 and 3000 hr, respectively. For the latter system, it appearsmore » feasible to operate fuel elements at central temperatures of approximates -2000 deg F to burnups in excess of 25,000 Mwd/MTU. Niobium and zirconium alloys have shown promise as cladding materials for the utilization of hyperstoichiometric uranium monocarbide. Out-of-pile tests have shown that zirconium and niobium alloys are compatible with hyperstoichiometric UC to 1600 deg F. Zirconium alloys are strength limited above 1150 deg F. Niobium, while satisfactory from a strength standpoint at 1600 deg F, may be limited by nuclear considerations. (auth)« less

Authors:
Publication Date:
Research Org.:
Atomics International. Div. of North American Aviation, Inc., Canoga Park, Calif.
OSTI Identifier:
4752823
Report Number(s):
NAA-SR-7696
NSA Number:
NSA-17-016650
DOE Contract Number:  
AT(11-1)-GEN-8
Resource Type:
Technical Report
Resource Relation:
Other Information: Orig. Receipt Date: 31-DEC-63
Country of Publication:
United States
Language:
English
Subject:
METALS, CERAMICS, AND OTHER MATERIALS; BURNUP; CARBIDES; DENSITY; DIFFUSION; EXPANSION; FISSION PRODUCTS; FUEL CANS; GASES; HEAT TRANSFER; HIGH TEMPERATURE; IRRADIATION; LIQUID METAL COOLANT; NIOBIUM ALLOYS; PERFORMANCE; SODIUM; STAINLESS STEEL- 304; STAINLESS STEELS; TEMPERATURE; TENSILE PROPERTIES; URANIUM CARBIDES; ZIRCONIUM ALLOYS

Citation Formats

Hahn, R D. A STUDY OF URANIUM CARBIDE AND CLADDING MATERIALS FOR HIGH-TEMPERATURE SODIUM-COOLED REACTORS. United States: N. p., 1963. Web.
Hahn, R D. A STUDY OF URANIUM CARBIDE AND CLADDING MATERIALS FOR HIGH-TEMPERATURE SODIUM-COOLED REACTORS. United States.
Hahn, R D. 1963. "A STUDY OF URANIUM CARBIDE AND CLADDING MATERIALS FOR HIGH-TEMPERATURE SODIUM-COOLED REACTORS". United States.
@article{osti_4752823,
title = {A STUDY OF URANIUM CARBIDE AND CLADDING MATERIALS FOR HIGH-TEMPERATURE SODIUM-COOLED REACTORS},
author = {Hahn, R D},
abstractNote = {Irradiation experiments were conducted to determine the irradiation performance characteristics of uranium monocarbide, exposed to high-temperature ( approximates 2000 deg F), high heat flux ( approximates -10/sup 6/ Btu/hr-ft/sup 2/), high-burnup ( approximates 25,000 Mwd/ MTU) operation. Hyperstoichiometric uranium carbide fuels have sustained maximum conditions of burnup and average central temperature of 24,000 and 10,000 Mwd/ MTU at 1550 and 2150 deg F, respectively. Hypostoichiometric uranium carbide fuels were exposed to conditions of 1800 deg F average central temperature for a burnup of approximates 7000 Mwd/ MTU. Fission gas release, in general, was low. Values for fuel dimensional increases were scattered, due to thermal or mechanical effects over the fuel column lengths, and could not be correlated with burnup or average fuel central temperatures. The fuel density decrease profile of the fuel length was similar to the burnup profile. Hyperstoichiometric UC/Na (or NaK) Type 304 stainless steel systems show excessive cladding carburization at fuel surface temperatures as low as 1000 deg F for an acceptable reactor residence time. Hypostoichiometric UC/Na Type 304 stainless steel systems are compatible at 1200 and 1600 deg F under isothermal conditions for at least 6400 and 3000 hr, respectively. For the latter system, it appears feasible to operate fuel elements at central temperatures of approximates -2000 deg F to burnups in excess of 25,000 Mwd/MTU. Niobium and zirconium alloys have shown promise as cladding materials for the utilization of hyperstoichiometric uranium monocarbide. Out-of-pile tests have shown that zirconium and niobium alloys are compatible with hyperstoichiometric UC to 1600 deg F. Zirconium alloys are strength limited above 1150 deg F. Niobium, while satisfactory from a strength standpoint at 1600 deg F, may be limited by nuclear considerations. (auth)},
doi = {},
url = {https://www.osti.gov/biblio/4752823}, journal = {},
number = ,
volume = ,
place = {United States},
year = {1963},
month = {3}
}

Technical Report:
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