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High-temperature deformation and rupture behavior of internally-pressurized Zircaloy-4 cladding in vacuum and steam enivronments. [LOCA conditions]

Description: The high-temperature diametral expansion and rupture behavior of Zircaloy-4 fuel-cladding tubes have been investigated in vacuum and steam environments under transient-heating conditions that are of interest in hypothetical loss-of-coolant accident situations in light-water reactors. The effects of internal pressure, heating rate, axial constraint, and localized temperature nonuniformities in the cladding on the maximum circumferential strain have been determined for burst temperatures between … more
Date: January 1, 1977
Creator: Chung, H. M.; Garde, A. M. & Kassner, T. F.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Assessment of Void Swelling in Austenitic Stainless Steel Core Internals.

Description: As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevan… more
Date: January 31, 2006
Creator: Chung, H. M. & Technology, Energy
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Irradiation-Assisted Stress Corrosion Cracking Behavior of Austenitic Stainless Steels Applicable to LWR Core Internals.

Description: This report summarizes work performed at Argonne National Laboratory on irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels that were irradiated in the Halden reactor in simulation of irradiation-induced degradation of boiling water reactor (BWR) core internal components. Slow-strain-rate tensile tests in BWR-like oxidizing water were conducted on 27 austenitic stainless steel alloys that were irradiated at 288 C in helium to 0.4, 1.3, and 3.0 dpa. Fractographi… more
Date: January 31, 2006
Creator: Chung, H. M.; Shack, W. J. & Technology, Energy
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Environmentally assisted cracking in Light Water Reactors: Semiannual report, October 1994--March 1995. Volume 20

Description: This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) from October 1994 to March 1995. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, (b) EAC of Alloy 600 and 690, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water with several di… more
Date: January 1996
Creator: Chung, H. M.; Chopra, O. K.; Gavenda, D. J.; Hins, A. G.; Kassner, T. F.; Ruther, W. E. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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