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Iodine-induced Stress Corrosion Cracking (I-SCC) Workshop

Abstract Booklet

Chalk River Laboratories | 2015 October 22 - 23

Workshop Committee

Program Chair, Bill Richmond Membership Chair, Susan Yatabe Technical Chair, Markus Piro Technical Chair, Mike Welland

A Review of Pellet-Clad Interaction ResearchMarkus H.A. Piro (CNL, Canada)

Pellet-Clad Interaction (PCI) in zirconium alloy cladding has been a long-standing concern in the context of nuclear fuel reliability. Mitigation measures have constrained operational procedures in combination with modifying the manufacturing process of the cladding, adding a composite layer or a protective coating to the inner surface of the cladding. The foregoing remedies have satisfactorily resolved the problem at the expense of restricted operational performance (i.e., maximum power, power ramping and burnup) that limit potential revenue in operation. Furthermore, the implementation of these measures does not preclude the reoccurrence of PCI failures following other design changes in the future. The objective of this review is to present the current state of understanding of the phenomenon, outline current mitigation strategies, and discuss recent advancements in experimental and computational techniques to further elucidate the problem.

Modern “State of the Art” SCC Testing With On-line Monitoring of Crack Initiation and Crack Growth – Application to I-SCC QuestionsMike Wright (CNL, Canada)

Methods to monitor and measure SCC in real time as cracks grow have revolutionized the way SCC (and environmentally assisted cracking in general) is studied and have advanced the science of SCC tremendously over the last 20 years. Measuring crack growth, in real time, and as a function of the fracture mechanics concept of crack-tip stress intensity were developed from the field of fatigue testing, by way of “corrosion” fatigue testing. A key aspects of this technology is the ability to control loading of the specimen in response to changes in crack length. For SCC testing, and for corrosion fatigue testing, this must all be achieved with the specimen within an appropriate environmental chamber or autoclave. For fully optimised test systems the environment is also monitored and controlled so as to maintain constant conditions, or manipulated during a test so as to explore crack growth rates changes in direct response to changes in the environment, while keeping all other variable constant. In the 1990s it was recognized that this technology represented more than a tool for generating crack growth rate data.

The ability to monitor and control SCC on-line in real time was exploited to explore and quantify the cracking mechanisms themselves by quantifying response to material, stress and environmental variables under precisely controlled conditions. The adoption of fracture mechanics based testing with on-line monitoring and control of test conditions has resulted in significant advances in the fundamental understanding of cracking mechanisms and their dependencies in the context of the manifestations of SCC and corrosion fatigue that affect austenitic alloys employed as pressure boundary, or structural components in PWRs and BWRS. For example, extensive (and prolonged) laboratory testing revealed that the concept of crack/no-crack thresholds is misleading. The response of the SCC has been found to be more a continuum, and not characterized by thresholds beyond which cracking stops. Also, these techniques have been successful in allowing the effects of cold work, elevated hardness associated with weld heat affected zones and irradiation damage, to be quantified and rationalized.

More recently the industry has started to adapt the crack growth rate experimental and testing methodology to investigate crack initiation. The key principles are the same. There is a requirement to measure and control the stress/strain conditions precisely and to monitor and measure crack initiation as a function of time or “cycles” whether mechanical loading cycles or environmental exposure cycles. The key is to measure meaningful cracking behaviour as a function of a controlled variable while keeping all other variables constant.

The techniques that are now the norm for investigating SCC of austenitic alloys in reactor coolant environments have not been applied in the context of I-SCC (fuel sheathing SCC). There are two reasons for this. To date, there has been no pressing need to do so and the fuel sheathing itself is thin walled; too thin to extract useful fracture mechanics specimens from. This latter point need not be a barrier if representative material of sufficient thickness is available. Considering, some of the unresolved questions over I-SCC dependencies, it is expected that fracture mechanics based test techniques could address some of these very effectively.

Will This System Fail by Stress Corrosion Cracking?R. Winston Revie (Consultant, Canada)

Predicting the occurrence of stress corrosion cracking using accelerated tests will be reviewed, focusing on environmental, mechanical, and metallurgical test conditions. Controlling these conditions in small-scale experiments to be similar to the realistic engineering conditions in the system being assessed for stress corrosion cracking will be discussed, in order to avoid misleading indicators and “prediction traps”. Based on the results of predictive tests, options for managing stress corrosion cracking can be developed, ensuring enhanced reliability and safety over the anticipated engineering lifetime.

Application of State of the Art Analytical TEM to SCC ResearchSuraj Persaud (CNL, Canada)

The last 15 years have seen major advances in the capabilities of analytical transmission electron microscopy (TEM). In parallel, the development of sample extraction using “focused ion beam” (FIB) technology has greatly facilitated the application of analytical TEM to studies of SCC. The use of FIB sample extraction has solved the historical problem of how to extract a small TEM sample from precisely the location of interest. In most cases this region of interest is the very tip of a growing crack. Today it is relatively easy to obtain TEM foils from crack tips without compromising the sample. Modern analytical TEMs can then be used to characterise material and chemistry of crack tips. The presentation will illustrate how these techniques are applied and what can be learned. Alternatively, the region of interest might be a surface/oxide interface or a grain boundary. The combination of FIB sample extraction combined with modern analytical TEM is particularly attractive and effective when examining irradiated material. Almost all the hands-on aspects of sample preparation are completely avoided and the technique is highly material efficient. Some very recent developments in the field will be presented. Although CNL does not have its own facilities for these examinations we do work closely with the Universities and National Labs around the world who are at the cutting edge applying the techniques to SCC problems in the nuclear industry.

Early History of Power Ramp Failures and Potential MitigationsKit Coleman (CNL, Canada)

In the 1960s some failures of Zircaloy-clad UO2 fuel were associated with power ramps after some burnup. Two main causes were postulated: overloading from fuel expansion and corrosion by fission products. The two effects could be combined as stress corrosion. The tensile stress was exacerbated by interaction with radial cracks in the fuel and the potential corroding elements were identified as iodine, cesium and cadmium. Mitigations were based on reducing the tensile stresses and preventing the corrodants from reaching the inside surface of the Zircaloy cladding. A thin graphite coating was used on the inside surface of CANDU fuel cladding while pure zirconium (plus a small amount of Fe) was used on the inside surface of fuel cladding in Boiling Water Reactors. Despite the success of these methods, the mechanism of the cracking process is still under discussion and is the subject of this workshop.

A Summary of CNL SCC Related IrradiationsNoel Harrison (CNL, Canada), Stavros Corbett (CNL, Canada), Lori Walters (CNL, Canada)

CNL has a long history of fuel and materials irradiations, much of it conducted in the NRX and NRU research reactors. While the research has covered many areas of nuclear fuel and materials design, Stress Corrosion Cracking identification and mitigation strategies have been predominant. The presentation will identify the irradiation facilities used, and outline recent experiments focussed on Stress Corrosion Cracking, as well as related topics such as establishing thresholds and protective sheath coatings.

Post Irradiation Examination of Liquid Metal Bond Hydride Fuel in Zirconium CladdingAndrew M. Casella (PNNL, USA), David J. Senor (PNNL, USA), Edgar C. Buck (PNNL, USA), Donald R. Olander (UC Berkeley, USA), Kurt A. Terrani (UC Berkeley, USA), Peter Hosemann (UC Berkeley, USA), Mehdi Balooch (UC Berkeley, USA)

A radiation experiment on a novel fuel form was conducted at the MIT reactor up to 0.29%FIMA on U0.17 ZrH1.6 fuel. Pellets were manufactured and filled in Zircalloy-2 tubes. The gap was back-filled with lead Bismuth Eutectic (LBE) before reactor irradiation. The rodlet was encapsulated in a sealed canister, back-filled with the same LBE. The fuel center-line and rodlet outside temperature was monitored during the irradiation exposure. At 0.29%FIMA the rodlet was taken out of the MIT reactor and characterized in cross sections using optical and scanning electron microscopy. It was found that the liquid metal reduced the cladding wall thickness while the fuel seems largely intact. Detailed data as well as potential explanations for the cladding loss is given in this presentation.

Pellet-Associated Cladding Degradation (PACE): A collaborative research consortium to investigate Pellet Cladding Interaction (PCI)Philipp Fränkel (U. Manchester, UK)

PCI is clearly a very complex phenomenon, requiring improved understanding of a number of interdependent material behaviours. Combining the strengths and capabilities of a range of institutions interested in this area with current industry knowledge and experience will provide a strong platform to achieving this goal. The overriding aim of the PACE consortium is to coordinate research activities in order to provide a more mechanistic understanding of the PCI phenomena. The talk will give an overview of the PACE consortium and highlight current and planned investigations at Manchester as part of PACE. We aim to combine state-of-the-art experimental techniques and proton irradiation experiments with atomistic simulation to gain a new understanding of the underlying mechanisms of PCI.

Use of Ion-Beam Irradiation Facilities for the Study of Corrosion Mark Daymond (Queen’s U., Canada)

Microchemistry of Fuel-Sheath Interfaces by X-ray Photoelectron SpectroscopyThan Do (CNL, Canada), William H. Hocking (CNL, Canada), Ian J. Muir (CNL, Canada)

An advanced facility for characterization of highly radioactive materials by Imaging X-ray Photoelectron Spectroscopy (XPS) has been developed at the Chalk River Laboratories, based on more than a decade of prior experience with a prototype system. The system is equipped with auxiliary electron and ion guns to provide additional in situ capabilities for scanning electron microscopy (SEM), scanning Auger microscopy (SAM) and composition depth profiling. The application of this facility to characterize irradiated fuel materials will be illustrated with selected results taken from detailed study of the microchemistry at the fuel-sheath interfaces from irradiated fuel elements. Inside surfaces of the end caps and the welds between the sheath and the end caps were analyzed as well as various locations on the thin-walled Zircaloy-4 sheath. The in situ SEM capability was essential for selecting different areas on each sample, such as sheath locations with and without a visible retained graphite layer, for XPS analysis. Effective infiltration of segregated fission products, into the graphite was demonstrated by depth profiling. A richer chemistry of segregated fission products was found on the end caps than on the sheath, with elevated levels of barium, strontium, tellurium, iodine and cadmium, and cesium. The results are consistent with current understanding of the primary migration route for fission products to the sheath and also indicate that the graphite layer functions as a chemical barrier to segregated fission products.

In-situ Neutron Diffraction Study of Zircaloy-4 Subjected to Biaxial TensionMichael A Gharghouri (CNL, Canada), Daniel McDonald (CNL, Canada), Lin Xiao (CNL, Canada)

Deformation Behaviour of Zircaloy-4 Subjected to Uniaxial and Biaxial Loadingin Air and Iodine EnvironmentLin Xiao (CNL, Canada), Rosaura Ham-Su (CNL, Canada), Daniel McDonald, (CNL, Canada), Michael Gharghouri (CNBC, Canada)

Zircaloy-4 cladding commonly undergoes complex biaxial deformation resulting from the pellet-cladding mechanical interaction due to fuel thermal expansion during power transients and the internal pressure of the fission gases; consequently, stress corrosion cracking (SCC) may take place. Understanding SCC is fundamental for the design of new nuclear fuel cycles and for increasing fuel burnup. The objective of this work is to study the deformation behaviour of Zircaloy-4 in air and iodine environments, and ascertain the deformation and damage mechanisms during uniaxial and biaxial loading so as to understand the fundamental mechanism of SCC. The effect of iodine solution and displacement rate on uniaxial tensile behaviour of Zircaloy-4 was studied. The results reveal that displacement rate has a significant effect on the tensile strength and ductility of Zircaloy-4 in iodine-methanol solution when the iodine concentration is 1.0%. The effect diminishes as the iodine concentration decreases. The effect of iodine on strength can be ignored; however, the ductility still significantly decreases, when the iodine concentration is less than 0.001%. Macroscopic equivalent stress-strain curves reveal that Zircaloy-4 under biaxial loading at the stress ratio of 1.0 has the highest yield stress among different stress ratios. Biaxial deformation substructure was analyzed with transmission electron microscope. After deformation at the stress ratio of 0.5, the predominant feature is parallel dislocation lines produced by {1010} prismatic slip. As the stress ratio increases to 1.0, the dislocation configuration changes from parallel dislocation lines to embryonic dislocation cells. The activated slip system changes from predominant {1010} prismatic slip at the stress ratio 0.5 to {1010} prismatic slip, {1011} pyramidal slip and {1012} twinning, simultaneously. Multiple systems slip and twin at the stress ratio 1.0 mitigate the anisotropy of plastic deformation.

Mitigating the Iodine-induced Stress Corrosion Cracking of Zircaloy-4 Fuel Sheathing using Siloxane CoatingsGraham A. Ferrier (RMC, Canada), M. (Kommy) Farahani (RMC, Canada), Paul K. Chan (RMC, Canada), andEmily C. Corcoran (RMC, Canada)

For approximately fifty years, a thin (3-20 μm) graphite-based coating known as CANLUB has played an important role in limiting the iodine-induced 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·kgU-1). Consequently, commercially-available siloxane coatings are evaluated by their 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·cm-3), mechanical deflection tests and scanning electron microscopy show that a particular siloxane coating outperforms the graphite coating in preserving the mechanical integrity of the rings. Note that the dipping process of the siloxane coating is comparable with the graphite coating process. The resiliency of this siloxane coating was further confirmed by subjecting coated Zircaloy-4 tubes to mechanical vibration, forced adhesion, crush tests, scratch tests, and a 50-day exposure to thermal neutron flux ((2.5±0.1)x1011 n·cm-2·s-1) in the SLOWPOKE-2 nuclear reactor at the Royal Military College of Canada.

Pellet cladding mechanical interaction and thermally induced fracture modeling with BISONKyle Gamble (INL, USA), Benjamin Spencer (INL, USA)

BISON is a multidimensional multiphysics fuel performance code under development at the Idaho National Laboratory. It is based on a multiphysics finite element framework, which permits it to naturally support a variety of fuel types, including LWR fuel rods, TRISO fuel particles, and metallic fuel. Recent work with BISON has demonstrated its abilities to use 3D models to represent the effects of local irregularities in the fuel geometry on pellet-clad mechanical interaction (PCMI). BISON captures the local elevated temperatures, stresses, and strains in the cladding in the vicinity of the irregularity in the fuel. In addition, advanced models for fracture in the fuel have been developed based on the extended finite element method (XFEM) and peridynamic methods in this multiphysics environment.

Both the capabilities for modeling PCMI and those for modeling fuel fracture in BISON will be useful for future work modeling Iodine-induced stress corrosion cracking (SCC) in cladding. Stress concentrations in the cladding are important contributors to SCC, and can be caused by local defects in pellets and cracks. A natural next step in the fracture modeling in BISON will be to use the capabilities in development to represent the interaction between discrete cracks and cladding. In addition, these same capabilities can potentially be used to model crack propagation in the cladding either at the macro or micro scale. This talk will provide a summary of the current work in 3D PCMI modeling and fracture modeling in BISON, with a discussion on the next steps in using these capabilities together for modeling of Iodine-induced SCC.

Finite Element Modelling of Fuel Sheath Strains and Stresses in Support of a Stress Corrosion Cracking ModelAnthony F. Williams (CNL, Canada) and Susan Yatabe (CNL, Canada)

As the name suggests, the phenomenon of stress corrosion cracking requires both a concentrated stress and the presence of a corrosive element such as iodine or cesium. The formation of stress corrosion cracks occurs at locations of stress concentration in the fuel sheathing or cladding due to mechanical interaction of the fuel pellet and the sheath/cladding. The development of a successful model must combine an understanding of both the stress formation and the chemistry of the corrosive element that results in crack formation and propagation. Over the past few years, we have been developing detailed 3D thermomechanical models of CANDU fuel pins using the ANSYS finite element software package. These finite element models have the potential for calculating in great detail the strains and stresses in the fuel sheathing that occur due to the pellet-sheath interaction. In particular these models include the phenomenon of pellet hourglassing and the corresponding stress concentrations that occur in the sheath at the pellet-to-pellet interfaces. These detailed calculations of the stress concentrations may act as the basis for the mechanical part of a future SCC model.

Recent Developments of the FAST Fuel ModelAndrew Prudil (CNL, Canada), Paul K. Chan (RMC, Canada), Brent J. Lewis (UOIT, Canada)

The 2D fuel performance code FAST has been enhanced to improve modeling of Pellet-Cladding Mechanical Interaction (PCMI) by incorporating models for fuel creep, time dependent cladding plasticity and fuel end caps. In addition, the separate models of grain boundary fission gas swelling, saturation and release kinetics have been replaced with a single mechanistic model of grain boundary fission gas behavior. The code is being used to model a Nuclear Energy Agency Expert Group on Reactor Fuel Performance (NEA-EGRFP) benchmark on PCMI. A 3D version of the code is being developed to investigate element bending due to non-axially symmetric mechanical loading in Pressurized Heavy-Water Reactor (PHWR) fuel (i.e., gravity, contact, and refueling operations) or temperature profiles due to coolant flow patterns or neutron flux gradients. Separate versions of the code have also been developed for thoria based mixed-oxide fuels and for modeling iodine induced stress corrosion cracking.

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