Defect testing of coextruded uranium-zircaloy-II clad fuel material in a 300 C out-of-reactor recirculating water loop: Interim report

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A major problem in the development of a pressurized water reactor coolant system for the NPS is the rupture performance of the fuel elements. As water temperatures are increased to 300 C, uranium corrosion rates increase rapidly. Swelling of the uranium fuel by corrosion could cause the process tube to burst or reduce the tube cooling water flow below acceptable limits. The desirability of slow cooling of the water to avoid thermal shocks to the reactor piping after a rupture is detected further complicates discharge and decontamination problems as fuel will continue to corrode with attendant fuel element damage during … continued below

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32 p.

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Hayden, K. D. & Goffard, J. W. September 25, 1959.

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  • Hanford Atomic Products Operation
    Publisher Info: General Electric Co., Richland, WA (United States). Hanford Atomic Products Operation
    Place of Publication: Richland, Washington

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Description

A major problem in the development of a pressurized water reactor coolant system for the NPS is the rupture performance of the fuel elements. As water temperatures are increased to 300 C, uranium corrosion rates increase rapidly. Swelling of the uranium fuel by corrosion could cause the process tube to burst or reduce the tube cooling water flow below acceptable limits. The desirability of slow cooling of the water to avoid thermal shocks to the reactor piping after a rupture is detected further complicates discharge and decontamination problems as fuel will continue to corrode with attendant fuel element damage during the cooling period. Coextruded uranium-zircaloy-2 clad fuel elements are scheduled for use in the NPR. The rupture behavior of this type fuel material after heat treatment was studied in ELMO-4, an out-of-reactor recirculating water loop. Several types of initial defects were studied. Fuel materials with five different heat treatment histories and with different types of defects were tested to determine their rupture behaviors. The five conditions were (1) as-extruded material as received from Nuclear Metals, Inc., (2) beta heat-treated and water quenched, (3) beta heat-treated and air-cooled, (4) beta heat-treated, isothermally treated at 600 C and air-cooled, and (5) beta heat-treated and furnace cooled (vacuum). Both pinhole and cleavage type defects were studied. Most of the work consisted of measuring weight loss and physical dimension changes after various lengths of exposure in 300 C water. This report presents the results of the tests.

Physical Description

32 p.

Notes

OSTI as DE94015693; Paper copy available at OSTI: phone, 865-576-8401, or email, reports@adonis.osti.gov

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  • Other Information: PBD: 25 Sep 1959

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Office of Scientific & Technical Information Technical Reports

Reports, articles and other documents harvested from the Office of Scientific and Technical Information.

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  • September 25, 1959

Added to The UNT Digital Library

  • Nov. 28, 2018, 2:33 p.m.

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  • Dec. 13, 2018, 3:23 p.m.

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Hayden, K. D. & Goffard, J. W. Defect testing of coextruded uranium-zircaloy-II clad fuel material in a 300 C out-of-reactor recirculating water loop: Interim report, report, September 25, 1959; Richland, Washington. (https://digital.library.unt.edu/ark:/67531/metadc1338846/: accessed February 13, 2025), University of North Texas Libraries, UNT Digital Library, https://digital.library.unt.edu; crediting UNT Libraries Government Documents Department.

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