Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “STRESS CORROSION”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 811 records · Page 45

Welding Process Development for Spent Nuclear Fuel Canister Repair

The potential for stress corrosion cracking (SCC) of welded stainless-steel interim storage containers for spent nuclear fuel (SNF) has been identified as a high priority data gap. This paper presents a fusion welding process that was developed for SNF canister repair. Submerged arc welding (SAW) was developed to weld 12.7 mm (0.5 in.) thick 304L stainless steel plates to simulate the initial welds on SNF canisters. The SAW procedure was qualified following ASME Boiler and Pressure Vessel Code requirements. During SAW, the welding temperature was recorded at various locations by using thermocouples. After SAW, weld microstructures were characterized, joint mechanical properties were tested, and the maximum tensile residual stress direction was identified. After SAW procedure qualification, artificial cracks were excavated perpendicular to the maximum tensile residual stress direction in the SAW heat affected zone. Machine cold-wire gas tungsten arc welding (CW-GTAW) was developed and used for repair welding at cracked locations.

Tang, Wei↗

Modeling of Stress Distribution in Molten Salt Thermal Energy Storage Tanks for In-Service Central Receiver Power Plants

Failures in molten nitrate salts thermal energy storage tanks (TES) have been occurring in several concentrating solar power (CSP) plants around the world after a few months or years of operation. These failures are mainly related to a combination of high stress, corrosion, large deformation, and thermal cycling. Most of these failures are a consequence of the infancy of this technology; technical standards for design, fabrication, implementation, and operation of molten salt tanks operating at temperatures up to 565 degrees Celsius do not exist and most of current in-service tanks have been designed using the recommendations from the American Petroleum Institute (API) 650 and the American Society of Mechanical Engineers ASME section II standards, which are applicable for TES tanks operating at lower temperatures. We have developed a 3-dimensional model of a representative molten salt hot tank design that uses typical CSP plant operation conditions and allows determining the temperature and stress distribution in the tank. The model also considers the initial floor residual stresses and distortion after its welding fabrication. We have identified critical thermal gradients in the tank floor and shell and locations in the tank where the stresses are high and could become susceptible points for failure.

concentrating solar power↗

Development of flaw acceptance criteria for aging management of spent nuclear fuel multiple-purpose canisters

A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. The canister may be subject to service-induced degradation when it is exposed to aggressive atmospheric environments during a possibly long-term storage period if the permanent repository is yet to be identified and readied. Because heat treatment for stress relief is not required for the construction of an MPC, stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic In-service Inspection. The first-order instability flaw sizes has been determined with bounding flaw configurations, that is, through-wall axial or circumferential cracks, and part-through-wall long axial flaw or 360° circumferential crack. The procedure recommended by the American Petroleum Institute (API) 579 Fitness-for-Service code (Second Edition) is used to estimate the instability crack length or depth by implementing the failure assessment diagram (FAD) methodology. The welding residual stresses are mostly unknown and are therefore estimated with the API 579 procedure. It is demonstrated in this paper that the residual stress has significant impact on the instability length or depth of the crack. The findings will limit the applicability of the flaw tolerance obtained from limit load approach where residual stress is ignored and only ligament yielding is considered.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Analysis of Localized Deformation Processes in Highly Irradiated Austenitic Stainless Steel through In Situ Techniques

This report describes new experimental results obtained during in situ mechanical tests with neutron-irradiated (10.7 dpa) austenitic 304L steel specimens. The in situ tensile tests inside a scanning electron microscope (SEM) were accompanied by modern high-resolution electron backscatter diffraction (HR-EBSD) analysis. The HR-EBSD approach allowed for investigating and quantifying internal stresses and dislocation structures in nondeformed and deformed steels, materials used in light water reactor internal components. The literature regarding the HR-EBSD analysis of metals and alloys is very limited, and practically no publications on the HR-EBSD analysis coupled with in situ tests are available. Experimental work, performed at Oak Ridge National Laboratory’s Low Activation Materials Development and Analysis facility, is detailed in this report. Section 1 briefly describes strain and stress localization processes and underlines their importance and role in irradiated materials, materials used in nuclear power plants. The section demonstrates that very limited quantitative data exist regarding localized stresses and strains in the irradiated materials at the meso-level (i.e., inside the grain, for a group of interacting grains, or for channel–grain boundary [GB] interaction) whereas local stresses at GBs are key contributors to the stress corrosion crack initiation.Section 2 documents the key methods used to conduct the experimental work: the SEM/EBSD system, miniature tensile frame, and irradiated material. Irradiated specimen manufacturing is described, as this option may be of special interest for other projects. Section 3 provides details of the HR-EBSD in situ tests (e.g., tensile curves, geometry, dimensions of regions of interest, load, and deformation steps). Section 4 discusses the results related to the evolution of geometrically necessary dislocations (GNDs) and their density, spatial, and statistical distributions.Section 5 focuses on the HR-EBSD analysis with respect to a channel–GB interaction. Two aspects are discussed in detail: mapping and analyzing the GND fields before and after loading, and the possibility of investigating local stresses and elastic strains at the channel–GB interaction. The present work results are beneficial for understanding strain and stress localization processes leading to crack initiation, for analyzing the material after long-term in-service life, and for developing predictive models. Section 6 summarizes the work performed and discusses future activities within this research direction.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Survey of baffle plate samples, specimen machining, and hydrogen measurements

This report outlines preliminary results on the survey and characterization of irradiated baffle plate, baffle former, and flux thimble tube specimens as part of the ongoing investigation into irradiation-induced embrittlement in austenitic stainless steels at ambient (room temperature) conditions. The specimens originated from commercial pressurized water reactor (PWR) components, covering displacement damage doses ranging from approximately 0.065 to 75 dpa. Initial scanning electron microscopy (SEM) surveys revealed that specimen surfaces exhibited fine machining marks and in-service-formed oxide layers on the side surfaces of the analyzed specimens. The oxide layers revealed specific features resembling localized "pitting-like" corrosion. An additional set of specimens representing in-service oxidation is being preserved for future microstructure analysis work. As believed, this will provide additional insights into long-term material degradation in PRWs. Specimen machining challenges emerged due to a complex failure of the electric discharge machining (EDM) system located in hot area. Despite partial restoration, issues persist with the EDM’s secondary power supply, necessitating the exploration of alternative EDM or computer numerical control (CNC) machining approaches to facilitate tensile specimen preparation. Currently, low speed saw cutting is in progress to prepare specimens for hydrogen measurements and general microstructure analysis. The near-term goals include completing tensile specimen machining for advanced mechanical testing and characterizing fracture mechanisms, and stress-corrosion cracking testing, ultimately aiming to identify and mitigate the ambient-condition intergranular cracking through targeted post-irradiation annealing strategies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Cryogenic Fracture Toughness Improvement for the Super Lightweight Tank's Main Structural Alloy

Marshall Space Flight Center has developed a two-step (TS) artificial aging technique that can significantly enhance cryogenic fracture toughness and resistance to stress corrosion cracking (SCC) in aluminum-copper-lithium alloy 2195. The new TS aging treatment consists of exposures at 132 C (270 F)/20 hr + 138 C (280 F)/42 hr, which can be readily applied to flight hardware production. TS aging achieves the same yield strength levels as conventional aging, while providing much improved ductility in the short transverse direction. After TS aging, five previously rejected lots of alloy 2195 (lots 950M029B, 960M030F, 960M030J, 960M030K, and 960M030L) passed simulated service testing for use in the super lightweight tank program. Each lot exhibited higher fracture toughness at cryogenic temperature than at ambient temperature. Their SCC resistance was also enhanced. All SCC specimens passed the minimum 10-day requirement in 3.5-percent sodium chloride alternate immersion at a stress of 45 ksi. The SCC lives ranged from 57 to 83 days, with an average of 70 days.

Chen, P. S.↗

A FEASIBILITY STUDY ON CRACK REPAIR IN AUSTENITIC STAINLESS STEEL DRY STORAGE CANISTERS USING ISOTHERMAL FRICTION STIR WELDING

In recent time, solid state crack repair techniques in spent nuclear fuel dry storage canisters (SNF-DSC) have garnered significant interest. Plates of austenitic stainless steels are usually arc welded to fabricate the cylindrical canisters which are prone to chloride-induced stress corrosion cracking. Friction stir welding (FSW) is considered to be a potential candidate for damage repair. In this work, electrical discharge machining was used to simulate cracks in 304L stainless steel plates of 12.7 mm thickness. Subsequently, isothermal FSW was carried out at two different temperatures (725 and 825°C) in order to repair the simulated crack. Microstructure-mechanical property correlations in FSWed plates were studied in detail. Significant grain refinement observed in the stir zone (SZ) was attributed to dynamic recrystallization occurring during FSW. A comparison of Vickers microhardness showed higher hardness and homogeneity in the 725°C SZ than its 825°C counterpart. The fraction of S3 boundary is found to be low in the SZ as compared to base metal (BM). Yield strength of the joints was found to be approximately 100 MPa higher than that of the BM accompanied by a drop in ductility by a factor of 2. Residual stress profile across 825°C weld was measured using x-ray diffraction.

Bhattacharyya, Madhumanti↗

Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimating clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. In conclusion, based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

"Ultra"-Fast Fracture Strength of Advanced Structural Ceramic Materials Studied at Elevated Temperatures

The accurate determination of inert strength is important in reliable life prediction of structural ceramic components. At ambient temperature, the inert strength of a brittle material is typically regarded as free of the effects of slow crack growth due to stress corrosion. Therefore, the inert strength can be determined either by eliminating active species, especially moisture, with an appropriate inert medium, or by using a very high test rate. However, at elevated temperatures, the concept or definition of the inert strength of brittle ceramic materials is not clear, since temperature itself is a degrading environment, resulting in strength degradation through slow crack growth and/or creep. Since the mechanism to control strength is rate-dependent viscous flow, the only conceivable way to determine the inert strength at elevated temperatures is to utilize a very fast test rate that either minimizes the time for or eliminates slow crack growth. Few experimental studies have measured the elevated-temperature, inert (or "ultra"-fast fracture) strength of advanced ceramics. At the NASA Lewis Research Center, an experimental study was initiated to better understand the "ultra"-fast fracture strength behavior of advanced ceramics at elevated temperatures. Fourteen advanced ceramics - one alumina, eleven silicon nitrides, and two silicon carbides - have been tested using constant stress-rate (dynamic fatigue) testing in flexure with a series of stress rates including the "ultra"-fast stress rate of 33 000 MPa/sec with digitally controlled test frames. The results for these 14 advanced ceramics indicate that, notwithstanding possible changes in flaw populations as well as flaw configurations because of elevated temperatures, the strength at 33 000 MPa/sec approached the room-temperature strength or reached a higher value than that determined at the conventional test rate of 30 MPa/sec. On the basis of the experimental data, it can be stated that the elevated-temperature, inert strength of an advanced ceramic material can be defined as the strength where no slow crack growth takes place at the temperature.

Choi, Sung R.↗

Analysis of Base Metal Microstructures and Mechanical Properties of a Single-Piece, Spin-Formed Forward Pressure Vessel Bulkhead

In 2012, a pathfinder forward pressure vessel bulkhead (FPVBH) for the Orion Multi-Purpose Crew Vehicle (MPCV) was fabricated using aluminum (Al) alloy 2219. The demonstration article was offered as a replacement for the baseline configuration, which at that time was a multi-piece welded construction using aluminum-lithium (Al-Li) alloy 2195. In 2014, the NASA Engineering and Safety Center (NESC) funded a Phase I effort to explore spin-forming as a manufacturing method to produce a complex-shaped, single-piece FPVBH for the crew module (CM) [1]. The primary focus of the NESC effort was to expand spin-forming technology to the fabrication of an Al-Li 2195 FPVBH. As a result, limited testing and analysis was performed on the original Al 2219 article. In this study, a sufficiently large plate of Al 2219 was not available for the fabrication of a FPVBH. Therefore, the forming blank comprised two plates (from different lots) butt welded together using a single friction stir weld. The welded forming blank was then spin formed into the FPVBH configuration and heat treated to the T62 temper. Mechanical property testing of the spin-formed 2219-T62 FPVBH revealed that the tensile properties were comparable to the Metallic Material Properties Development and Standardization (MMPDS) A-basis allowable for Al 2219-T6 wrought products. However, post-spin forming application of the standard heat treatment produced large, recrystallized grains in the material. The response varied within the two plates; one plate lot exhibited nearly complete recrystallization, while the other lot contained bands of recrystallization. Although no reduction in tensile properties was observed due to these microstructural variations, a more detailed investigation was recommended to better understand the effect of spin forming on the recrystallization mechanism and the impact on other material properties. The current study examined the microstructure and mechanical properties of the fully processed Al 2219-T62 FPVBH with emphasis on comparing differences between the two plates. Testing was limited to the cone region of the FPVBH. The two specific tasks in this study were: a) Characterization of the microstructure and mechanical properties in the two plates used in the FPVBH. b) Isolation of the impact of each step in the thermo-mechanical processing (TMP) sequence of the FPVBH on recrystallization and grain growth. Mechanical property testing sampled regions containing both large, recrystallized and small, unrecrystallized grains. Tensile strengths and elongations were equivalent in both regions and exceeded MMPDS A-basis allowables. Fracture toughness was equivalent in both regions and fracture morphology was typical of aluminum alloys. The fatigue precrack region of the fracture toughness specimens appeared more faceted in the recrystallized plate. This suggests that property testing that is more sensitive to grain size, such as fatigue crack growth rate and stress corrosion cracking, may be warranted. Alcoa reviewed the production records for the two plate lots and confirmed that composition of both plates was within the allowable range for Al 2219 and that plate processing steps were all performed within production limits. However, the Fe content was notably higher in the fully recrystallized plate, resulting in a greater volume fraction of large Fe-bearing inclusions and Al-Cu constituent particles. Alcoa did not provide details regarding plate processing variables for the two plate lots. It was surmised that deformation by hot rolling was the most accurate TMP simulation of commercial spin forming within a laboratory setting. Both hot rolling and spin forming comprise combinations of shear and compressive deformation. The through-thickness strain caused by hot rolling is symmetrical about the mid-plane, as the rollers contact both sides of the material. Shear strains are maximum at the inner and outer surfaces and trend toward zero at the mid-thickness. In contrast, spin forming utilizes a single roller on the outer surface, producing a gradient in shear stress through the thickness. Shear strains in spin-formed material are highest at the outer surface due to direct contact with the roller, and trend toward zero at the inner surface. Consequently, there will undoubtedly be some discrepancies in the mechanical response of hot-rolled and spin-formed products in the through-thickness direction. Interrupted TMP simulations were performed on remnant plate from the FPVBH forming blank in order to isolate the effects of a post-weld anneal, spin-forming deformation/thermal cycling, and solution heat treatment (SHT) on development of the recrystallized microstructures. Analysis confirmed that large, recrystallized grains formed exclusively during SHT and only after a certain deformation level was exceeded. Remnant plate from both lots that received thermal processing only did not exhibit recrystallization but exhibited fully recrystallized microstructures when processed by hot rolling. This indicates that (1) the level of deformation exceeded the critical level for recrystallization in both plates and (2) that the level of deformation during hot rolling was greater than that which accumulated during spin forming of the FPVBH. Comparison of the hot-rolled plate with the spin-formed material showed that the recrystallized grain size was larger in the FPVBH, providing further evidence that deformation levels were lower in the FPVBH. The most plausible explanation for the differing degrees of recrystallization in the two plates is particle stimulated nucleation (PSN) at the Fe-bearing inclusions and large Al-Cu constituent particles. In the plate with higher solute content, the greater volume fraction of these particles reduces the deformation level required to promote recrystallization. During spin forming the critical deformation level was likely exceeded for the higher Fe content plate resulting in a fully recrystallized microstructure. The lower Fe content plate developed a microstructure that exhibited bands of recrystallization due to the combination of a non-uniform distribution of particles and deformation. The critical deformation required for recrystallization was likely only exceeded during spin forming in regions of higher particle content. While there were differences in Fe content between these lots of Al 2219 plate, the composition of both was within defined alloy limits and Alcoa reported no anomalies during plate production. The observed differences in recrystallization after spin forming may reflect the inherent variability between these two lots of plate. The spin-forming vendor, Spincraft, reported that similar variations in recrystallized microstructures have been observed during examination of multiple spin-forming trials of a different 2xxx series aluminum alloy. Hence, tighter chemical specification limits on impurity levels in Al 2219 plate may help ensure uniform, predictable microstructures. Finally, the tensile and fracture toughness properties were not affected by the microstructural variations noted.

single-piece↗

Irradiation-assisted cracking of SA508-304L weldments with 308L groove filler and 309L butter in hot water immersion constant rate extension tests

Low alloy SA508 bainitic steel is often welded to 300 series austenitic stainless steel in nuclear power plants. These dissimilar metal weldments use 309L butter applied to the low alloy steel and 308L groove filler material to join the 309L to austenite. High alloy steels (such as the 308L and 309L material used here) have been identified as having greater susceptibility to irradiation-assisted stress corrosion cracking (IASCC). Evidence is presented here of cracking under tensile loads in 309L weldment butter in LWR hydrothermal conditions that correlates irradiation induced changes at δ ferrite–γ austenite phase boundaries and γ–γ grain boundaries; specifically intergranular Cr 23 C 6 precipitates. Cracking is not observed in irradiated 308L groove filler under the same conditions. Both the 309L butter and 308L filler material have a duplex or mixed δ–γ phase microstructure due to solidification during the welding procedure. We have performed constant extension rate autoclave immersion tests in BWR normal water chemistry hot water (288 °C, 10 MPa, 2000 wt. ppb dissolved oxygen, <100 nS/cm conductivity, neutral pH) of proton irradiated (2 MeV, approximately 3 displacements per atom at a depth of 10 μm) tensile specimens. Proton irradiation to the dpa levels used here induced Cr 23 C 6 carbide precipitation at γ–γ grain boundaries and δ–γ phase boundaries in the 309L butter but not the 308L filler. Cracking was observed in the proton-irradiated volume of 309L at these interfaces in the vicinity of the Cr 23 C 6 precipitates. We hypothesis the SA508 is the source of carbon in the 309L butter. Cr depletion at γ–γ grain boundaries was also observed and may also be a root cause of cracking. Furthermore, the 309L butter is under significant tensile residual stress in the weldment that adds to applied tensile loads and this too may be a contributing factor.

309L weld butter↗

Sulfur and Moisture Effects on Alumina Scale and TBC Spallation

It has been well established that a few ppmw sulfur impurity may segregate to the interface of thermally grown alumina scales and the underlying substrate, resulting in bond degradation and premature spallation. This has been shown for NiAl and NiCrAl-based alloys, bare single crystal superalloys, or coated superalloys. The role of reactive elements (especially Y) has been to getter the sulfur in the bulk and preclude interfacial segregation. Pt additions are also very beneficial, however a similar thermodynamic explanation does not apply. The purpose of the present discussion is to highlight some observations of these effects on Rene'142, Rene'N5, PWA1480, and PWA1484. For PWA1480, we have mapped cyclic oxidation and spallation in terms of potential sulfur interfacial layers and found that a cumulative amount of about one monolayer is sufficient to degrade long term adhesion. Depending on substrate thickness, optimum performance occurs if sulfur is reduced below about 0.2-0.5 ppmw. This is accomplished in the laboratory by hydrogen annealing or commercially by melt-fluxing. Excellent 1150 C cyclic oxidation is thus demonstrated for desulfurized Rene'142, Rene'N5, and PWA1484. Alternatively, a series of N5 alloys provided by GE-AE have shown that as little as 15 ppmw of Y dopant was effective in providing remarkable scale adhesion. In support of a Y-S gettering mechanism, hydrogen annealing was unable to desulfurize these alloys from their initial level of 5 ppmw S. This impurity and critical doping level corresponds closely to YS or Y2S3 stoichiometry. In many cases, Y-doped alloys or alloys with marginal sulfur levels exhibit an oxidative sensitivity to the ambient humidity called Moisture-Induced Delayed Spallation (MIDS). After substantial scale growth, coupled with damage from repeated cycling, cold samples may spall after a period of time, breathing on them, or immersing them in water. While stress corrosion arguments may apply, we propose that the underlying cause is related to a hydrogen embrittlement reaction: Al alloy + 3 H2O = Al(OH)3 + 3H(+) + 3e(-). This mechanism is derived from an analogous moisture-induced hydrogen embrittlement mechanism originally shown for Ni3Al and FeAl intermetallics. Consequently, a cathodic hydrogen charging technique was used to demonstrate that electrolytic de-scaling occurs for these otherwise adherent alumina scales formed on Y-doped Rene'N5, in support of hydrogen effects. Finally, some TBC observations are discussed in light of all of the above. Plasma sprayed 8YSZ coatings, produced on PWA1484 without a bond coat, were found to survive more than 1000 1-hr cycles at 1100 C when desulfurized to below 0.1 ppmw. At higher sulfur (1.2 ppmw) levels, moisture sensitivity and delayed TBC failure, referred to as Desk Top Spallation, occurred at just 200 hr. Despite a large degree of scatter, a factor of 5 in life improvement is indicated for desulfurized samples in cyclic furnace tests, confirming the beneficial effect of low sulfur alloys on model TBC systems. (DTS and moisture effects are also observed on commercially applied PVD 7YSZ coatings on Rene'N5+Y with Pt-aluminide bond coats). These types of catastrophic failure were subverted on the model system by segmenting the substrate into a network of 0.010 high ribs, spaced in. apart, prior to plasma spraying. No failures occurred after 1000 cycles at 1150 C or after 2000 cycles at 1100 C, even after water immersion. The benefit is described in terms of elasticity models and a critical buckling stress.

Smialek, James L.↗

Role of Grain Boundary Cr5B3 Precipitates on Intergranular Attack in Alloy 600

It is well established that thermally treated (TT) alloy 600 exhibits superior resistance to intergranular stress corrosion cracking (IGSCC) in pressurized water reactor (PWR) primary water environments than do solution annealed (SA) and mill annealed (MA) equivalents of the same material. This improved resistance is nominally ascribed to the prevalence of grain boundary (GB) Cr-carbide precipitates (M7C3 or M23C6). In this study, we perform high-resolution characterization by scanning transmission electron microscopy (STEM) and atom probe tomography (APT) on a heat of alloy 600 with a very low concentration of C (0.01 at.%) and a modest B concentration (46 appm). After SA+TT annealing, this heat exhibits IG Cr-boride precipitates (Cr5B3) in the absence of Cr carbides. Despite IG precipitation of Cr5B3, the pristine GB shows no measurable Cr depletion and B remains segregated at the GB. During intergranular attack (IGA) in PWR primary water at 360 °C, the Cr5B3 precipitates dissolve rapidly. The observed IGA is subtly different from what is typical for alloy 600, specifically with higher concentrations of Li and B within the corrosion oxides while the GB ahead exhibits less depletion of oxidizing species (e.g. Cr, Fe, Si and B) than other alloy 600 heats. Together these observations suggest that IG precipitation of Cr5B3 in the absence of Cr carbides has a neutral to slightly positive effect on the IG corrosion resistance of alloy 600 GBs in PWR primary water.

alloy 600, APT, TEM↗

Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial (R-Z) or plane radial-circumferential (R-θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used in this paper to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. Finally, in comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Grain boundary engineering for structure materials of nuclear reactors

Grain boundary engineering (GBE), primarily implemented by thermomechanical processing, is an effective and economical method of enhancing the properties of polycrystalline materials. Among the factors affecting grain boundary character distribution, literature data showed definitive effect of grain size and texture. GBE is more effective for austenitic stainless steels and Ni-base alloys compared to other structural materials of nuclear reactors, such as refractory metals, ferritic and ferritic–martensitic steels, and Zr alloys. Furthermore, GBE has shown beneficial effects on improving the strength, creep strength, and resistance to stress corrosion cracking and oxidation of austenitic stainless steels and Ni-base alloys.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Oxidation and Hydrogen Uptake in Zirconium, Zircaloy-2 and Zircaloy-4: Computational Thermodynamics and Ab Initio Calculations

Zirconium-based alloys Zircaloy-2 and Zircaloy-4 are widely used in nuclear industry as cladding materials for BWRs and PWRs, respectively. Over more than 60 years these materials displayed a very good combination of properties such as low neutron absorption, creep behavior, stress-corrosion cracking resistance, reduced hydrogen uptake, corrosion, and/or oxidation, especially in the case of Zircaloy-4. However, over the last couple of years energetic efforts were undertaken to improve their oxidation resistance during off-normal temperature excursions, as well as to further improve upon the already achieved levels of mechanical behavior and reduced hydrogen uptake. In order to facilitate the development of such novel materials, it is very important to achieve not only engineering control, but also scientific understanding of the underlying material degradation mechanisms, both in working conditions and in storage of spent nuclear fuel. This paper strives to contribute to these efforts by constructing the thermodynamic models of both alloys, constructing of the respective phase diagrams, and oxidation mechanisms. A special emphasis was placed upon the role of zirconium suboxides in hydrogen uptake reduction and the atomic mechanisms of oxidation. To that end, computational thermodynamics calculations were conducted concurrently with first-principles atomistic modeling.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Microstructural characterization of cold-worked 316 stainless steel flux thimble tubes irradiated up to 100 dpa in a commercial Pressurized Water Reactor

Two flux thimble tubes (FTT) made of 15% cold-worked 316 stainless steel (SS) were harvested from Ringhals Pressurized Water Reactor (PWR) Unit 2, with peak damages of 76 and 100 displacements per atom (dpa) after 29 and 34 years’ service, respectively. Specimens sectioned from parent tubes were comprehensively characterized with nominal damage levels of ~0, ~41, ~74, 76, and 100 dpa at a nominal temperature range of 285-323 °C. Both FTTs contained helium and hydrogen gases as transmutation products. The helium follows a production rate of ~9.8 appm/dpa, while environmental factors complicate hydrogen production obscuring an exact H/dpa ratio. Irradiation-induced dislocation loops, nano-cavities, solute clusters, and microsegregation were all observed. The dislocation loops and nano-cavities indicated saturation at 41 dpa. The solute clusters continued to evolve with Ni-Si clusters formed at 41 dpa, and Ni-Si-Mn-P clusters formed at 74 and 100 dpa, but neither clusters exhibited distinct diffraction patterns at any damage levels. Solute clusters were observed to frequently be co-located with dislocation loops, but fully decorated loops were rarely detected. Significant radiation-induced segregation (RIS) was observed around grain boundaries at all damage levels. The modified inverse Kirkendall (MIK) model captured the RIS behavior of major elements. Large cavities within or around an Mn-S rich region were observed for the first time. Through all the damage levels, void swelling is always below 0.05%, making significant dimensional change unlikely in core internals when used at similar conditions. Meanwhile, the role of overwhelming nanocavities, presumably helium bubbles, should be considered in other potential degradation mechanisms, including irradiation-assisted stress corrosion cracking, embrittlement, and loss of fracture toughness, which remain the concerns for extended operation of nuclear power plants.

Post Irradiation Examination, Isotope Dilution Mas↗