The effect of sulfur content on weld penetration in austenitic stainless steel orbital tube welds
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Materials science↗
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Microstructural Evolution During a Solid State Tube Pinch Weld of Type 304l Stainless Steel
Microstructure development is examined for a specialized spot weld that is used as a solid-state closure process for austenitic stainless steel tubing, referred to as pinch welding. In order to elucidate the microstructural evolution of the weld, a series of test welds were made at nominal conditions using tubing and production like components. These pinch welds normally terminate after twelve cycles of a 60 Hz AC weld process. In this study, production tubes were welded from one to twelve cycles and the microstructure and weld variables after each individual weld cycle number were characterized using radiography and optical metallography. Two electrochemical etchants were used to highlight different microstructural features. The study revealed that: (1) this type pinch weld is largely complete after about six cycles of 60 Hz AC current, half the weld time utilized; (2) the resistance, deformation, and closure length approach “steady-state” conditions after six cycles; and (3) both oxalic and nitric acid electrolytic etchants are useful for highlighting specific microstructural attributes of type 304 L stainless steel. Finally, two distinct microstructural regions can be identified for these welds: the edge of the weld, which is driven by concentrated deformation, recrystallization, and grain growth, and the center region, which is more typical of forge welding and micro-asperity breakdown followed by diffusion and grain-growth.
36 MATERIALS SCIENCE↗
Diffuser Braze Development
At the request of SRTE, through the PDRD project, SRNL is developing a vacuum brazing process to facilitate the fabrication of hydrogen purification systems. In particular, this effort is directed at tube-style hydrogen diffusers. The diffusers are comprised of numerous palladium-silver tubes that are brazed to a nickel fitting, which is subsequently welded to a stainless steel tube, fitting, or tube sheet. Historically, torch brazing has been used as a heat source; for this work vacuum furnace brazing was suggested to improve the uniformity of the process. This report describes the braze process development for braze selection and simple nickel fitting to tube brazing. Five braze alloys were selected and sessile drop testing was conducted over a range of temperatures from 25 to 75° C over the liquidus temperature. Samples were characterized for wetting angle, erosion, and brazability. The cup-to-tube braze joints were inspected using digital radiography, ultrasonic testing, and metallographic examination. Two of the five alloys exhibited excellent fabrication properties and are considered for further development.
36 MATERIALS SCIENCE↗
DETERMINATION OF THE SOLID-STATE RESISTANCE-WELDABILITY OF ADDITIVELY MANUFACTURED 304L STAINLESS STEEL
Additive manufacturing of pressure-containing boundaries is of interest to numerous industries, so a project was initiated to develop several different methods to join and weld additively manufactured materials. The approach investigated in this study was a low energy, solid state spot weld. This general approach has been routinely used for conventionally prepared tubing and is now being expanded to additively manufactured components. Additively manufactured 304L stainless steel sheets, semi-tubes, and tubes were resistance spot welded and compared to comparable welds in conventionally manufactured tubing. Radiographic computed tomography and metallography were used to evaluate the suitability of the welds using three criteria: weld closure length, thickness, and width. All additively manufactured test welds closely matched the conventional tubing for these criteria
additive manufacturing↗
Fracture toughness evaluations for spent nuclear fuel dry storage canister welds and spent nuclear fuel clad-pellet structures
Integrity of spent nuclear fuel (SNF) interim storage canisters is very important to the safety of the back-end nuclear fuel cycle. Stress corrosion cracking (SCC) potential of interim storage canister has been considered as a high priority. Because no post-weld heat treatment was required for forming these canisters, the high tensile residual stress existed within these canister welds. This can change the fracture resistance capacity significantly as well as increase SCC potential. Due to relative thin shell thickness of a canister weldment, the spiral notch torsion test (SNTT) method was used to estimate the canister weldment fracture toughness. SNTT was developed to measure the intrinsic fracture toughness of structural materials using small specimens. The SNTT method has been applied to a wide variety of structural materials, such as low-alloy steels, stainless steel, aluminum alloy, ceramics, concrete, and composites. The SNTT system operates by applying pure torsion to cylindrical specimens with a notch line that spirals around the specimen at a 45° pitch. In order to carry out pure torsion load mode, biaxial tension/torsion tester was developed accordingly to perform SNTT protocol. Moreover, applying fracture mechanics approach to SNF system reliability investigation is warranted due to the inherent flaws and hydride structures existed in a SNF system after nuclear reactor operation. However, none of the existing fracture toughness data deal with fuel cladding specific geometry or spent fuel material conditions, such as cladding structure with the pellet-inserts and the associated pellet clad mechanical interactions induced mixed-mode damage mechanisms. Thus, the development of an intrinsic fracture mechanics approach that is suitable for SNF materials is needed. Furthermore, due to thin wall and small dimension of clad tubing structure, the SNTT method was used to estimate the clad tubing structure fracture toughness. Fracture testing were performed on the received stainless steel canister weldment, most SNTT weld samples fracture initiation sites are at heat-affected zone (HAZ) regions. The estimated fracture toughness J Q ’ for the baseline SS304 steel is at 283 kJ/m². The estimated JQ’ for the SS304/308 weld and baseline metals are 148 kJ/m 2 and 459 kJ/m 2 , respectively. Out of cell fracture testing for spent fuel structure were carried out on the surrogate rods made of Zr-4 clad and alumina inserts, the estimated fracture toughness values for baseline Zr-4 cladding with alumina-pellet inserts are: (1) For SNTT samples with a short or medium crack length, between 5.4-mm and 8-mm, the estimate J IQ upon fracture initiation for the baseline Zr-4 cladding is at 50 kJ/m 2 with 2-sigma uncertainty of 3.26 kJ/m 2 , and the associated K IQ is at 67.46 MPa$\sqrt{m}$; and (2) For SNTT samples with a long crack length, around 13-mm, the crack initialization is deviated from that of the Mode-I tensile fracture and appears to be a mixed-mode fracture of Mode I - tensile stress and Mode III - out of plane shear stress; the estimated J MQ is at 18.9 kJ/m 2 , the associated K MQ is at 41.4 MPa$\sqrt{m}$.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Determination of the solid state resistance weldability of additively manufactured 304L stainless steel
PURPOSE What? • Determine the weldability of additively manufactured (AM) 304L Stainless Steel ( • Why? • Use AM to make gas containing vessels that will be seal welded • How? • Prepare • Weld • Compare mock tubes to SS tubing
Rogers, Jeremy K.↗
Radionuclide tracing based in situ corrosion and mass transport monitoring of 316L stainless steel in a molten salt closed loop
In the study, we report an in situ corrosion and mass transport monitoring method developed using a radionuclide tracing technique for the corrosion study of 316L stainless steel (316L SS) in a NaCl–MgCl 2 eutectic molten salt natural circulation loop. This method involves cyclotron irradiation of a small tube section with 16 MeV protons, later welds at the hot leg of the molten salt flow loop, generating radionuclides 51 Cr, 52 Mn, and 56 Co at the salt–alloy interface. By measuring the activity variations of these radionuclides at different sections along the loop, both the in situ monitoring of the corrosion attack depth of 316L SS and corrosion product transport and its precipitation in flowing NaCl–MgCl 2 molten salt are achieved. While 316L SS is the focus of this study, the technique reported herein can be extended to other structural materials being used in a wide range of industrial applications.
36 MATERIALS SCIENCE↗
Characterization of Long-term Service Coal Combustion Power Plant Extreme Environment Materials
The technical goal of this project is to develop a comprehensive database of mechanical properties, alloy microstructures, and to a lesser extent, the oxidation/corrosion behavior of coal-fired power plant components, such as boiler tubing, steam headers and steam piping, which have been in long-term service (at least 100,000 operating hours, and preferably more than 200,000 operating hours) under the operating conditions of high temperatures and high mechanical stresses where creep, fatigue, steam-side oxidation, and fireside corrosion are life-limiting factors. The components included in the database will include the following types of high temperature alloys: creep strength enhanced ferritic (CSEF) steels, 300-series H-grade stainless steels, and dissimilar metal welds (DMWs).
Characterization↗
Determination of the Solid-State Resistance-Weldability of Additively Manufactured 304L Stainless Steel
• A scoping study was undertaken at Savannah River National Laboratory (SRNL) to determine the solidstate resistance-weldability of additively manufactured 304L stainless steel. • Additive manufacturing of pressure-containing boundaries is of interest to numerous industries. • The approach investigated in this study was a low energy, solid state spot weld (pinch weld). • This general approach has been routinely used for conventionally prepared tubing and is now being expanded to additively manufactured components. • This work aimed to understand the impact of AM on the bond quality and weld geometry of pinch welds.
Rogers, Jeremy K. [Savannah River National Laborat↗
Characterization of Long-Term Service Coal Combustion Power Plant Extreme Environment Materials
The objective of this DOE-sponsored project was to develop a comprehensive database of mechanical properties, alloy microstructures, and to a lesser extent, the oxidation/corrosion behaviors of coal-fired power plant components, such as boiler tubing, steam headers, and steam piping, which had been in service for at least 100,000 operating hours (preferably more than 200,000 operating hours) under the operating conditions of high temperatures and high mechanical stresses where creep, fatigue, steam-side oxidation, and fireside corrosion were life-limiting factors. The components included in this database consisted of ferritic steels, creep strength enhanced ferritic (CSEF) steels, and 300-series H-grade stainless steels, as well as dissimilar metal welds (DMWs) among these types of materials. As a result of extensive metallurgical characterization and mechanical testing performed in this project, a comprehensive database on mechanical properties and detailed quantitative microstructural information was successfully developed for several long-term serviced EEM components. Such a database can be used by material research communities to develop, calibrate, refine, and validate mechanical behaviors, models, and other assessment tools for accurate prediction of remaining life of major components under similar EEM operating conditions.
20 FOSSIL-FUELED POWER PLANTS↗
Test and Validate Distributed Coaxial Cable Sensors for in situ Condition Monitoring of Coal-Fired Boiler Tubes
This project aims to test, validate, and advance the technology readiness level (from TRL5 to TRL7) of a novel low-cost distributed stainless-steel/ceramic coaxial cable sensing (SSC-CCS) technology for in situ monitoring of the boiler tube temperature in existing coal-fired power plants. The novel SSC-CCS sensing technology and associated condition-based monitoring (CBM) software to be demonstrated in this project will lead to an improved understanding of the boiler tube failure mechanisms and a prognostic system to improve the overall performance, reliability, and flexibility of the nation’s coal-fired power plant fleet. A boiler tube monitoring system with distributed coaxial cable temperature sensors and a sensor acquisition system was constructed. The high-temperature coaxial cable sensor with a length of 1.3m was made by using a quartz tube (1mm inner diameter (ID) and 6mm outer diameter (OD)) to concentrically separate a 304 stainless-steel (SS) rod (1mm OD) and SS tube (7.94mm OD and 6.16mm ID). The sensor acquisition system includes a vector network analyzer (VNA), a radio frequency (RF) power amplifier, multiple switches and a USB hub. The distributed stainless-steel quartz coaxial cable sensor (SSQ-CCS) had a linear response to temperature with a resolution uncertainty of σ = 0.77℃. To withstand the harsh conditions of 3,300 steam pressures and 800℃ high temperatures, the sensor was shielded by a protective tube made of the same material as the boiler tube. The protection tube had an OD of 1.5 inches and a thickness of 0.25 inches. In the laboratory tests, the sensor showed good sensitivity and fast response. The drift was bounded between +0.33% and -0.67% during a test at 600℃ for 350 hours, indicating good stability of the sensor. A field test was conducted where four sensors were welded on four superheat tubes (SH-Ts) at a coal-fired power station over 400 days. Conventional thermocouples were welded to the superheater tubes alongside the coaxial cable sensors for the purpose of comparison. Two sensors were capable of distributed sensing, with three multiplexed sensing sections. The other two sensors were single section. During the 400-day test period, the power plant experienced startups and shutdowns. At the steady state operations, the temperature of the boiler tube is about 600℃ (1112°F). The sensors recorded the entire coal-firing processes (start-up, steady state, and shut-down) and the glitch event. A GSM modem and a Watchdog were added to the system to ensure reliable data recording. The GSM modem sent daily messages to plant managers and Clemson team to inform the status of the sensor system. If the system was not normally working, the Watchdog would reboot the system automatically. The new coaxial cable based distributed sensing technology has been proven to be successful in both laboratory and field tests. A comprehensive four-stage multi-physics computational framework has been developed to assist the design, optimization, installation, and operation of SSQ-CCS. With the consideration of various operation conditions, we predict the distributions of flue gas temperatures within coal-fired boilers, the temperature correlation between the boiler tube and SSQ-CCS, and the safety of SSQ-CCS. A conditional-based monitoring system is implemented as well. The computational framework developed in this work can guide the future operation of coal-fired plants and other power plants for the safety prediction of boiler operations.
01 COAL, LIGNITE, AND PEAT↗
2019 accomplishments: degradation of mechanical properties in structural metals and welds for FTS reservoirs
This report documents work in 2019 at SRNL in support of the Aging and Lifetimes program. Specimens for fracture toughness testing were prepared in prior years by thermally pre-charging samples, provided by Sandia National Laboratories, of welded types 304L and 21-6-9 stainless steels. These samples were pre-charged in 2017. The fracture toughness of the first aged specimens was measured and compared to historical data. The measurements of the 304L and 21-6-9 weld and particularly HAZ represent some of the first fracture data relevant to modern GTS reservoirs. This work represents an on-going collaboration between SRNL and SNL to understand tritium embrittlement of structural metals in Gas Transfer System reservoirs which informs lifetime assessments. Continued testing was conducted on ongoing historical studies, including the effects of very long-term aging, and the relative performance of differently forged stockpile relevant 304L and 316L samples. New samples were pre-charged with tritium and placed in a freezer for an aging study to allow helium to build into the material. The precharged samples consist of smooth and notched tensile samples of 304L stainless steel, as well as tube tensile samples of different thermal treatments supplied by Sandia National Laboratories. Their charging marks the beginning of a new sequence of studies to explore fundamental deformation mechanisms underpinning tritium embrittlement. At predesignated times, samples will be removed from aging and tested to provide data which will help develop an understanding of material degradation with increasing aging time. Also documented is progress in employing various advanced characterization techniques including high resolution electron microscopy, thermally programed desorption, and small angle x-ray scattering, as well as facilities renovations and equipment procurements in support of sustained tritium impacts on materials program at SRNL.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
2021 Report - SRNL Aging and Lifetimes program tritium aging studies on structural alloys
This report documents work performed in fiscal year 2021 at SRNL in support of the Aging and Lifetimes program. This work is part of an on-going collaboration between SRNL and SNL to understand tritium embrittlement of structural metals used in Gas Transfer System reservoirs, thereby informing lifetime assessments and predictions. In this effort, test coupons are pre-charged with tritium and allowed to age in a freezer to minimize tritium diffusion and off-gassing while allowing for born-in helium levels to systematically increase. Coupons are then removed periodically and tested to develop an understanding of how mechanical properties degrade as helium levels increase. This report summarizes test results from coupons which are in various stages of aging as they were tritium pre-charged in previous years. Tests were performed on 304L that was previously tritium pre-charged and allowed to age to nominally 300 appm helium. Specimens from this study were previously tested at 100 appm nominal helium content, and this new test point expands that study. Tensile testing of these heat treated 304L tubes (and welded tubes) aged to 100 appm helium were completed which provided a unique study of a diverse set of microstructures and yield strengths ranging from 200-800 MPa. Additionally, 304L notched and smooth tensile specimens aged to 300 appm helium were tensile tested to predetermined strains to provide for a systematic microscopy investigation of aging effects on microstructural damage in partnership with SNL. Testing was performed on 304L electron beam (EB) weld coupons as well as 21-6-9 annealed coupons that were thermally pre-charged in 2020. Tests on the EB welds were completed at 100 appm helium contents, and tests on the annealed 21-6-9 material were completed at 100, 200, and 300 appm helium. Table 0-1 provides the status of the ongoing tritium coupon aging studies. The table includes the year tritium pre-charging took place, the material, and the targeted helium concentrations (appm) for testing.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Characterization and prediction of the electromechanical wear of contact tips during wire arc additive manufacturing of 316L stainless steel
Here, this study seeks to better understand the degradation of the contact tip with respect to WAAM for a 316L wire electrode as well as explore methods of monitoring the contact tip state from process data. The contact tip, a consumable component, positions the wire and serves as the electrical contact surface between the wire electrode and the welding power supply. The wear of the contact tip was characterized in terms of material loss and material contamination for a set of tips worn to discrete levels as measured by the amount of wire fed or arc time. Geometrical characterization found a 49% increase in the bore exit area at 180 meters of wire fed. Machine learning models were developed to predict the relative bore exit area of the contact tip from arc-based process data and a random forest classifier exhibited favorable performance with a cross-validated f1-score of 0.84. The regression architecture implemented a multi-layer perceptron with the ability to predict the relative exit area with an $R^2$ score of 0.75. Key features used in the prediction include the standard deviation of the voltage and the time between shorts.
Contact tip wear↗
A Design and Fabrication Options Trade Study for Metallic Fuel without Internal Sodium Bonding
Nuclear fuels using alloys of uranium, or metallic fuels, have many beneficial properties. The classical metallic fuel design uses a loose fitting cylindrical “slug” of fuel placed inside stainless-steel cladding tubes where the gap is filled with sodium. This sodium bond is liquid at operating temperature and conducts heat from the slug to the cladding, especially in early life before fuel swells into contact with the cladding. Despite the benefits of sodium bonding, there is a desire to develop metallic fuel technologies without it chiefly to reduce chemical reaction hazards in spent fuel storage from sodium fast reactors operating on once-through fuel cycles. Elimination of the sodium bond may also help unlock potential benefits for fuel fabrication, reactor neutronics, and compatibility with other types of reactors. Creating a sodium-free metallic fuel revolves around the problem of manufacturing fuel slug geometries which are in close contact with the cladding at beginning of life to facilitate heat transport while alleviating fuel-cladding chemical interactions (FCCI) at this interface and providing enough free volume to accommodate fuel swelling. Accelerating development and qualification of this fuel system will require careful selection of design and manufacturing options. To this end, a design trade-off study was performed to evaluate candidate options. Several design and manufacturing options were assessed, weighted, scored, and ranked with respect to fabrication, normal reactor operation, off-normal scenarios, and back-end considerations. This effort was performed both for “baseline” needs, which represented a once-through fuel cycle at temperatures and burnups known to be viable for sodium-bonded metallic fuel, and for “enhanced” needs to represent opportunities for closed fuel cycles and/or more aggressive temperatures/burnups. The outcomes of this study prioritized a baseline technology using U-Zr alloy with additives to mitigate FCCI, produced in annular slug geometry by continuous casting, clad in austenitic stainless-steel alloy, and followed by a final step to swage the cladding down to close the gap. This study prioritized an enhanced fuel technology using U-Mo alloy, also produced by continuous casting into an annular geometry, followed by coating/plating with an FCCI barrier on the slug, again with a final step to swage the cladding diameter down using oxide dispersion strengthened steel. It was noted that development of the enhanced fuel technology would entail more risk, thus U-Zr alloy was put forth as a backup to U-Mo if challenges are encountered with FCCI barriers, and advanced ferritic/martensitic steels are put forth as a backup to oxide dispersion strengthened steels if swaging and welding are found unworkable.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗