Mitigating the Stress Corrosion Cracking of Zircaloy-4 Fuel Sheathing: Siloxane Coatings Revisited

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Graham A. Ferrier ◽  
Mohsen Farahani ◽  
Joseph Metzler ◽  
Paul K. Chan ◽  
Emily C. Corcoran
Keyword(s):  
Wear Resistance ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Reactor ◽  
Corrosion Cracking ◽  
Candu Reactors ◽  
Graphite Coating ◽  
Water Reactors ◽  
And Performance ◽  
Graphite Coatings

For more than 50 years, a thin (3–20 μm) graphite coating has played an important role in limiting the stress corrosion cracking (SCC) of Zircaloy-4 fuel sheathing in CANDU® nuclear reactors. Siloxane coatings, which were examined alongside graphite coatings in the early 1970s, demonstrated even better tolerance against power-ramp-induced SCC and exhibited better wear resistance than graphite coatings. Although siloxane technology developed significantly in the 1980s/1990s, siloxane coatings remain unused in CANDU reactors, because graphite is relatively inexpensive and performs well in-service. However, advanced CANDU designs will accommodate average burnups, exceeding the threshold tolerable by the graphite coating (450  MWh/kgHE). In addition, siloxane coatings may find applicability in pressurized and boiling water reactors, wherein the burnups are inherently larger than those in CANDU reactors. Consequently, a commercially available siloxane coating is evaluated by its present-day chemistry, wear resistance, and performance in hot, stressful, and corrosive environments. After subjecting slotted Zircaloy-4 rings to iodine concentrations exceeding the estimated in-reactor concentration (1  mg/cm3), mechanical deflection tests and scanning electron microscopy (SEM) show that the siloxane coating outperforms the graphite coating in preserving the mechanical integrity of the rings. Furthermore, the baked siloxane coating survived a 50-day exposure to thermal neutron flux ((2.5±0.1)×1011  n/cm2 s) in the SLOWPOKE-2 nuclear reactor at the Royal Military College of Canada.

Microstructure and Microchemistry of Inconel 600 STEAM Generator Tubing

2000 ◽  
Vol 6 (S2) ◽  
pp. 356-357
Author(s):  
V. Perovic ◽  
A. Perovic ◽  
G.C. Weatherly ◽  
A.M. Brennenstuhl
Keyword(s):  
Grain Boundary ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Analytical Electron Microscopy ◽  
Corrosion Cracking ◽  
Steam Generators ◽  
Intergranular Stress Corrosion ◽  
Candu Reactors ◽  
Inconel 600 ◽  
Water Reactors

Inconel 600 is an austenitic Ni-Cr-Fe alloy which is extensively used for tubing in steam generators of pressurized light water reactors (PWR) and CANDU heavy water reactors, because of its excellent mechanical properties and corrosion resistance. However, there have been instances of intergranular stress corrosion cracking of tubes in operating steam generators. The chemistry and the structure of grain boundaries and grain boundary precipitation have emerged as factors of prime importance in understanding stress corrosion cracking and intergranular attack of nickel-base alloys (see e.g. ref. l).In this study analytical electron microscopy was used to determine the microstructure of grain boundary and matrix precipitates, grain boundary chromium content and dislocation substructure of selected steam generating tubes of CANDU reactors. The results of the in-service materials are compared with as-received material. Two JEOL 2010 STEM instruments were used in this study.

scholarly journals POTENTIAL MITIGATION STRATEGIES FOR PREVENTING STRESS CORROSION CRACKING FAILURES IN HIGH-BURNUP CANDU FUEL

2018 ◽  
Vol 7 (2) ◽  
pp. 127-146 ◽  
Author(s):  
Markus Piro ◽  
Dion Sunderland ◽  
Winston Revie ◽  
Steve Livingstone ◽  
Ike Dimayuga ◽  
...  
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Lessons Learned ◽  
Mitigation Strategies ◽  
Nuclear Materials ◽  
Characterization Methods ◽  
Candu Reactors ◽  
High Degree ◽  
Fuel Doping

Potential mitigation strategies for preventing stress corrosion cracking (SCC) failures in CANDU fuel cladding that are based on lessons learned on both domestic and international fronts are discussed in this paper. Although SCC failures have not been a major concern in CANDU reactors in recent decades, they may resurface at higher burnup for conventional fuels or with nonconventional fuels that are currently being investigated, such as MOX or thoria-based fuels. The motivation of this work is to provide the foundation for considering possible remedies for SCC failures. Three candidate remedies are discussed, namely improved fabrication methods for fuel appendages, barrier-liner cladding, and fuel doping. In support of this effort, recent advances in experimental characterization methods are described—methods that have been successfully used in non-nuclear materials that can be used to further elucidate SCC behaviour in CANDU fuel. The overall objective is to outline a path forward for characterizing material behaviour as an essential part of investigating remedies to SCC failure. This will allow increased fuel discharge burnup, maximum linear power, and plant manoeuvrability, while maintaining a high degree of reliability.

Effects of Weld Overlays on Leak-Before-Break Margins

2011 ◽  
Author(s):  
G. Angah Miessi ◽  
Peter C. Riccardella ◽  
Peihua Jing
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Operating Conditions ◽  
Intergranular Stress Corrosion ◽  
Pressurized Water ◽  
Weld Overlay ◽  
Water Reactors ◽  
Weld Overlays ◽  
Leak Before Break

Weld overlays have been used to remedy intergranular stress corrosion cracking (IGSCC) in boiling water reactors (BWRs) since the 1980s. Overlays have also been applied in the last few years in pressurized water reactors (PWRs) where primary water stress corrosion cracking (PWSCC) has developed. The weld overlay provides a structural reinforcement with SCC resistant material and favorable residual stresses at the ID of the overlaid component. Leak-before-break (LBB) had been applied to several piping systems in PWRs prior to recognizing the PWSCC susceptibility of Alloy 82/182 welds. The application of the weld overlay changes the geometric configuration of the component and as such, the original LBB evaluation is updated to reflect the new configuration at the susceptible weld. This paper describes a generic leak-before-break (LBB) analysis program which demonstrates that the application of weld overlays always improves LBB margins, relative to un-overlaid, PWSCC susceptible welds when all the other parameters or variables of the analyses (loads, geometry, operating conditions, analysis method, etc…) are kept equal. Analyses are performed using LBB methodology previously approved by the US NRC for weld overlaid components. The analyses are performed for a range of nozzle sizes (from 6″ to 34″) spanning the nominal pipe sizes to which LBB has been commonly applied, using associated representative loads and operating conditions. The analyses are performed for both overlaid and un-overlaid configurations of the same nozzles, and using both fatigue and PWSCC crack morphologies in the leakage rate calculations and the LBB margins are compared to show the benefit of the weld overlays.

Stress corrosion cracking of the AISI 316L stainless steel HAZ in a PWR nuclear reactor environment

2011 ◽  
Vol 25 (1) ◽  
pp. 15-23 ◽  
Author(s):  
Mônica Maria de Abreu Mendonça Schvartzman ◽  
Marco Antônio Dutra Quinan ◽  
Wagner Reis da Costa Campos ◽  
Luciana Iglésias Lourenço Lima
Keyword(s):  
Stainless Steel ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
316L Stainless Steel ◽  
Nuclear Reactor ◽  
Corrosion Cracking ◽  
Aisi 316L ◽  
Aisi 316L Stainless Steel

Stress Corrosion Cracking of Cold Worked Austenitic Stainless Steels in Laboratory Primary PWR Environment

2004 ◽  
Author(s):  
Francois Vaillant ◽  
Thierry Couvant ◽  
Jean-Marie Boursier ◽  
Claude Amzallag ◽  
Yves Rouillon ◽  
...  
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Stainless Steels ◽  
Crack Depth ◽  
Austenitic Stainless Steels ◽  
Corrosion Cracking ◽  
Pressurized Water ◽  
Water Reactors ◽  
Cold Worked ◽  
Cracking Process

Austenitic Stainless Steels (ASS) are widespread in primary and auxiliary circuits of Pressurized Water Reactors (PWRs). Moreover, some components suffer stress corrosion cracking (SCC) under neutron irradiation. This degradation could be the result of the increase of hardness and / or the modification of chemical composition at the grain boundary by irradiation. In order to avoid complex and costly corrosion facilities, the effects of radiation hardening on the material are commonly simulated by applying a pre-strain on non-irradiated material prior to stress corrosion cracking tests. The typical features of the cracking process in primary environment at 360°C during CERTs included an initiation stage (composed of a true initiation time and a slow propagation regime leading to a crack depth lower than 50 μm), then a “rapid” propagation stage before mechanical failure. Pre-straining increased significantly CGRs and the mode of pre-straining could strongly modify the crack path. No significant cracking (< 50 μm) was obtained under a pure static loading. A dynamic loading (CERT or cyclic) was required and various thresholds (hardness, elongation, stress) for the occurrence of SCC were determined. An important R&D program is in progress to develop initiation and propagation models for SCC of austenitic SS in primary environment.

Effect of Contaminants on Resistance To Stress Corrosion Cracking Of Ni-Cr Alloy 600 in Pressurized Water

CORROSION ◽  
1965 ◽  
Vol 21 (1) ◽  
pp. 1-8 ◽  
Author(s):  
H. R. COPSON ◽  
S. W. DEAN
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Iron Alloy ◽  
Corrosion Cracking ◽  
Hydrocarbon Contamination ◽  
Alloy 600 ◽  
Pressurized Water ◽  
Lead Powder ◽  
Resistance To Stress ◽  
Water Reactors

Abstract Numerous tests have shown that Alloy 600, a 76 nickel-15 chromium-7 iron alloy, has excellent corrosion resist­ance in pressurized high temperature water. The present tests were undertaken to determine the influence of possible contaminants in 600 F (316 C) pressurized water on corrosion behavior, using both single and double U-bend specimens. The double U-bend provided a combined stress and crevice specimen. Contaminants were sodium fluoride, air, lead powder, lead oxide, a petroleum hydrocarbon, and a mixture of lead powder and the hydrocarbon. Contamination and aeration were much in excess of any condition likely to be encountered in pressurized water reactors. Under certain conditions, some contaminants induced stress corrosion cracking.

Technical Basis for Code Case N-894 Alternative Rules for Repair of Classes 1, 2, and 3 Austenitic Stainless Steel Piping With Thermal Fatigue Cracking

2020 ◽  
Author(s):  
Stephen Marlette ◽  
Steven L. McCraken ◽  
Christopher Lohse
Keyword(s):  
Stainless Steel ◽  
Stress Corrosion ◽  
Thermal Fatigue ◽  
Stress Corrosion Cracking ◽  
Fatigue Cracking ◽  
Corrosion Cracking ◽  
Weld Overlay ◽  
Water Reactors ◽  
Technical Basis ◽  
Weld Overlays

Abstract Since 1982 the nuclear industry has employed weld overlay repairs to address intergranular stress corrosion cracking (IGSCC) in boiling water reactors (BWR) and primary water stress corrosion cracking (PWSCC) in pressurized water reactors (PWR). The American Society of Mechanical Engineers (ASME) has created several documents to provide rules and guidelines for weld overlay repair of nuclear components that have experienced stress corrosion cracking (SCC). These documents include ASME Code Case N-504-4 and ASME Section XI, Nonmandatory Appendix Q which specifically address weld overlay repair of stainless steel components. Recently, stainless steel components that have experienced thermal fatigue cracking at the inner diameter surfaces have been repaired with structural weld overlays (SWOL) using the methodology of Code Cases N-504-4 and N-740-2. The SWOL is a good choice for repair of thermal fatigue cracks in piping because it provides structural reinforcement to the affected location and places the inside diameter (ID) surface into compression preventing, or significantly reducing, further flaw growth. However, the rules of Case N-504-4 and N-740 were not specifically written to address thermal fatigue cracking as the primary cause and may not adequately address design, analysis and examination requirements when thermal fatigue is the active mechanism because it is very different in nature than SCC. For example, SCC is driven by a combination of environment, steady state operating stresses, residual stresses from welding and fabrication processes, and operating temperature, whereas thermal fatigue is driven by thermal stress cycles resulting from fluid thermal cycling or stratification. The source of the thermal events that result in cracking may not be as well understood or predictable as SCC degradation. In addition, weld overlays applied to address SCC are constructed of SCC resistant material but are not resistant to thermal fatigue. Therefore, ASME Section XI recognized that alternative rules were needed for repair of piping damaged by thermal fatigue. This paper provides a technical basis for weld overlay repair of components that have experienced thermal fatigue cracking. It addresses design, analysis and examination requirements considering the nature of thermal fatigue in nuclear piping systems. The Code Case was originally drafted based on the industry accepted rules of Case N-504-4 and Appendix Q but includes appropriate modifications needed to address thermal fatigue cracking. These modifications include removing restrictions such as the delta ferrite limit for PWRs that is only applicable to address SCC in BWR environments, and enhancements to the examination requirements to ensure that the repaired location is adequately monitored throughout the remaining service life of the plant. The purpose of this paper is to document the technical basis for Code Case N-894, which is currently still under development by ASME Section XI.

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