This study examines the corrosion resistance of additively manufactured 316L stainless steel (SS) for nuclear applications across three environments: pressurized water reactor primary water (PWR PW), hot concentrated nitric acid, and seawater. Wire-feed laser additive manufacturing (WLAM) specimens showed oxidation behavior similar to wrought 316L SS in PWR PW, though stress corrosion cracking (SCC) susceptibility varied with heat treatment. In nitric acid testing, laser powder bed fusion (L-PBF) specimens demonstrated superior corrosion resistance compared to conventional SS, primarily due to improved intergranular corrosion resistance resulting from cleaner feedstock powder and rapid solidification rates that minimized grain boundary segregation. Laser metal deposition (LMD) repair studies in seawater environments successfully produced dense, crack-free repairs with good metallurgical bonding that matched the substrate’s mechanical properties while maintaining corrosion resistance. These results emphasize the importance of corrosion testing for additively manufactured components and understanding how their unique microstructures affect performance.

As part of a Department of Energy (DOE) funded program assessing advanced manufacturing techniques for Small Modular Reactor (SMR) applications, the Nuclear Advanced Manufacturing Research Centre (AMRC) and the Electric Power Research Institute (EPRI) have been developing Electron Beam Welding (EBW) parameters and procedures based upon SA508 Grade 3 Class 1 base material. The transition shell, a complex component connecting the lower assembly to the upper assembly is a shell that flares up with varying thicknesses across its section. The component due to its geometry could be built by near net shape powder metallurgy hot isostatic pressing instead of conventional forging techniques. The demonstrator transition shell here is built from several sub-forging as a training exercise. The complex geometry and joint configuration were selected to assess the EBW as a suitable technique. This paper presents results from the steady state welding in the 60-110 mm material thickness range, showing that weld properties meet specification requirements. Weld quality was assured by Time-of-Flight Diffraction (ToFD). The transition shell was completed by welding a flange to the assembly. The presented transition shell assembly represents 6 welded sections all fabricated in below 100 min total welding time.

Nuclear reactor inspections occasionally identify degraded materials in irradiated reactor components. Although mechanical repair options are possible, these repair solutions may be cost prohibitive or impractical to implement due to access restraints and/or the severity of the degradation. Welding repair of reactor components may input excessive heat into these irradiated materials resulting in diffusion of trace amounts of helium within the grain boundaries of the weld heat-affected zone (HAZ). Intergranular HAZ cracking can then result from the combination of this helium diffusion and high localized tensile stresses generated during weld cooling. It is therefore critical to characterize these zones and understand limitations for welding highly irradiated components to prevent helium-induced cracking. To accomplish this, typical reactor structural materials including Types 304L and 316L stainless steels and nickel-based Alloy 600/182 materials irradiated within the High Flux Isotope Reactor facility at Oak Ridge National Laboratory were used in this study for welding and evaluation. A phased array ultrasonic inspection system has been developed to characterize cracking in the weld samples. It provides remote controlled scanning and minimizes handling the samples, minimizing operator dose. The samples are inspected from the side opposite of the welds. The material and weld grain noise were evaluated at 10 MHz and found to be conducive to detecting cracking in the material and welds. Inspection of the samples comprises a 10 MHz phased array probe sweeping a focused longitudinal wave from -60° to 60° while the probe is raster scanned over the sample in small increments. The collected data is analyzed using UltraVision 3. Several of the irradiated samples were inspected prior to welding. Some of the samples had what appear to be small lamination defects in them. One irradiated welded sample has been tested to date with no cracking detected, which has been confirmed by destructive examination.

High-performance Ferritic (HiperFer) steels represent a promising materials innovation for next-generation thermal energy conversion systems, particularly in cyclically operating applications like concentrating solar thermal plants and heat storage power plants (Carnot batteries), where current market adoption is hindered by the lack of cost-effective, high-performance materials. HiperFer steels demonstrate superior fatigue resistance, creep strength, and corrosion resistance compared to conventional ferritic-martensitic 9-12 Cr steels and some austenitic stainless steels, making them potentially transformative for future energy technologies. This paper examines the microstructural mechanisms underlying HiperFer’s enhanced fatigue resistance in both short and long crack propagation, while also presenting current findings on salt corrosion properties and exploring potential alloying improvements for fusion reactor applications, highlighting the broad technical relevance of these innovative materials.

Extended storage of spent nuclear fuel (SNF) in intermediate dry cask storage systems (DCSS) due to lack of permanent repositories is one of the key issues for sustainability of the current domestic Light Water Reactor (LWR) fleet. The stainless steel canisters used for storage in DCSS are potentially susceptible to chloride-induced stress corrosion cracking (CISCC) due to a combination of tensile stresses, susceptible microstructure, and a corrosive chloride salt environment. This research assesses the viability of the cold-spray process as a solution to CISCC in DCSS when sprayed with miniature tooling within a characteristic confinement in two different capacities: cleaning and coating. In general, the cold-spray process uses pressurized and preheated inert gas to propel powders at supersonic velocities, while remaining solid-state. Cold-spray cleaning is an economical, non-deposition process that leverages the mechanical force of the propelled powders to remove corrosive buildup on the canister, whereas the cold spray coating process uses augmented parameters to deposit a coating for CISCC repair and mitigation purposes. Moreover, both processes have the potential to induce a surface compressive residual stress that is known to impede the initiation of CISCC. Surface morphology, deposition analysis, and microstructural developments in the near-surface region were examined. Additionally, cyclic corrosion testing (CCT) was conducted to elucidate the influence of cold-spray cleaning and coating on corrosion performance.

Miniature specimen tests are necessary to assess the mechanical properties of new fuel cladding alloys for next-generation nuclear reactors. The small specimen allows for extensive testing programs from limited volumes of material. However, there is a lack of testing equipment to perform high-temperature mechanical tests on the miniature specimen. This work presents the development of a high-temperature creep test system for miniature specimens with in situ scanning electron microscope (SEM) testing capability for real-time characterization. Here, we discuss the challenges of the development of the system, such as gripping the samples, loading, heating, cooling mechanisms, and strain measurement. The equipment is used to investigate the creep behavior of FeCrAl alloy Kanthal APMT, and the results are compared with conventional creep test data from the same batch of this material.

This presentation compares the corrosion resistance of uncoated Haynes 230 and SS316HS substrates to the same substrates coated with a Fe-based amorphous alloy. The substrates were exposed to highly corrosive media, FLiNaK, for 120 hours at 700 °C. The findings indicate that the thermal spray amorphous alloy coating provided superior corrosion resistance within the coatings while protecting the substrates against the aggressive environment. As a result, the new amorphous metal coating improved the substrate's lifespan by providing better protection against high-temperature corrosion, paving the way for a more efficient and cost-effective future in various industrial applications.

The incore instrumentation system of a pressurized water reactor (PWR) facilitates neutron flux mapping and temperature measurements at specific core locations. A guide conduit, extending from the seal table to the lower reactor pressure vessel head, guides and protects each incore guide thimble between the table and the lower reactor vessel head. Each flux thimble houses a detector and drive cable. Once filled with reactor coolant, the conduit becomes an extension of the reactor coolant pressure boundary. This paper reports the examination results of cracking detected in a TP304 stainless steel guide conduit adjacent to a fillet weld at the upper surface of a TP304 seal table. The cracking resulted in reactor coolant leakage that was detected by the presence of boric acid deposits on the exterior of the conduit and table. Failure analysis including dimensional measurements, chemical analysis, stereomicroscopy, metallography, and scanning electron microscopy showed that extensive cracking of the conduit and seal table material occurred due to stress corrosion cracking (SCC). Assessment showed that chlorine-containing deposits were present on the exterior of the conduit and on the surfaces of the seal table and were due to the design and operation of HVAC systems at the coastal plant. Stainless steels are susceptible to SCC in environments with elevated temperatures, chloride contents, and increased tensile stress – particularly in non-post weld heat treated (PWHT) weld regions and the heat affected zone (HAZ). This was the apparent primary cause of the failure. However, chloride-induced SCC of such materials typically results in transgranular crack propagation, whereas the observed cracks were indicative of intergranular stress corrosion cracking (IGSCC). Microstructural analysis showed that the observed cracks initiated in sensitized areas of material adjacent to the weld. Sensitization of the material caused chromium depletion from adjacent areas and increased susceptibility of the depleted areas to IGSCC. In this case, the most probable source of sensitization was related to welding and the long-term growth of grain boundary carbides nucleated during welding. This was considered a contributing cause to the failure.

This paper presents the CEN WS064 Prospective Group 2, a project involving different European stakeholders from more than 20 organizations with the objective to identify the needs and propose code developments research for the nuclear design and construction code RCC-MRx for innovative reactors with more onerous operational conditions: i) reactor components are generally exposed to higher temperatures; ii) have innovative and more corrosive coolants such as liquid lead or molten salt; iii) materials and components are generally exposed to higher radiation levels than light-water reactors. The main outputs of the CEN WS064 are code evolution proposals and proposals for pre-normative research in support of code evolution. The code evolution is driven by further improving safety and cost reduction. Nuclear Design Codes are robust engineering tools but should incorporate new technologies and research. The paper describes the adopted methodology and the rationale for identifying code evolution needs. Code evolution and research proposals will be discussed. Examples of proposals that will be discussed include: Guideline for design of material/components with innovative coolants, extension of design life to 60 years; qualification of new materials and components with advanced manufacturing. A general requirement is that code evolution and associated material and component qualification and codification need to be significantly accelerated for which new approaches such as AI tools will play an important role.

There is a critical lack of data on the mechanical behavior of candidate structural materials for advanced nuclear reactors under molten halide salt environments. Limited legacy data from the molten salt reactor experiment (MSRE) program showed a significant reduction in creep rupture strength of a Ni-base alloy in molten fluoride salt. With ongoing efforts to commercialize different molten salt reactor concepts, the industry can considerably benefit from quantitative information on the impact of molten halide salts on the engineering properties such as creep and fatigue strength of materials of interest. The present work aims to assess the role of molten salt corrosion on the creep behavior of three alloys 316H, 617 and 282 at 650-816 °C. Creep tests were conducted in fluoride (FLiNaK) and chloride (NaCl-MgCl 2 ) salts. Initial results from the ongoing testing will be presented which suggest that the molten salt environment caused a 25-50% reduction in creep rupture lifetime compared to air exposures. Physics-based corrosion and creep models were employed to gain some insights into the potential degradation mechanisms.

Laser additive manufacturing (AM) is being considered by the nuclear industry to manufacture net- shape components for advanced reactors and micro reactors. Part-to-part and vendor-to-vendor variations in part quality, microstructure, and mechanical properties are common for additively manufactured components, attributing to the different processing conditions. This work demonstrates the use of microstructurally graded specimen as a high throughput means to establish the relationship between process-microstructure-creep properties. Through graded specimen manufacturing, multiple microstructures, correlated to the processing conditions, can be produced in a single specimen. The effects of a solution annealing heat treatment on the microstructure and creep properties of AM 316H are investigated in this work. Using digital image correlation (DIC), the creep strain can be calculated in these graded regions, allowing for multiple microstructures to be probed in a single creep test. The solution annealing heat treatment was not sufficient in recrystallization of the large, elongated grains in the AM material; however, it was sufficient in removing the cellular structure commonly found in AM processed alloys creating a network of subgrains in their place. The resulting changes in microstructure and mechanical properties are presented. The heat treatment was found to generally increase the minimum creep rate, reduce the minimum creep rate, and reduce the ductility. Significant amounts of grain boundary carbides and cavitation were observed.

Local vacuum electron beam welding is an advanced manufacturing technology which has been investigated at Sheffield Forgemasters to develop as part of a cost-effective, reliable, agile, and robust manufacturing route for the next generation of civil nuclear reactors in the UK. A dedicated electron beam welding facility at Sheffield Forgemasters has been installed. This includes an x-ray enclosure, 100kW diode electron gun, 100T turntable, and weld parameter development vacuum chamber. A small modular reactor demonstrator vessel has successfully been manufactured with a wall thickness of 180 mm, including indication-free slope-in, steady- state and slope-out welding parameters. Electroslag strip cladding has also been investigated to demonstrate its viability in reactor pressure vessel manufacture. The electro-slag strip cladding method has been shown to produce high quality 60 mm strips on a 2600 mm inner diameter ring.

An attempt is being made to develop novel Ni-Mo-W-Cr-Al-X alloys with ICME approach with critical experimental/simulations and processing/microstructural characterization/property evaluation and performance testing has been adopted. In this work, based on thermodynamic modeling five alloy compositions with varying Mo/W and two alloys with high tungsten modified with the addition of Al or Ti were selected and prepared. The newly developed alloys were evaluated for their response to thermal aging in the temperature range of 700 to 850 °C and corrosion in the KCl-NaCl-MgCl 2 salt under suitable conditions. Thermally aged and post-corrosion test samples were characterized to ascertain phase transformations, microstructural changes and corrosion mechanisms. Al/Ti modified alloys showed significant change in hardness after 400 hours aging at 750°C, which was found to be due to the presence of fine γ’/γ” precipitates along with plate-shaped W/Mo-rich particles. These alloys show comparable molten salt corrosion resistance as commercial alloys at 750°C for 200-hour exposures. The good corrosion behavior of these alloys may be attributed to the formation of a protective multicomponent Al-or Ti-enriched oxide as well as the unique microstructure.

A significant research and development effort is underway to support the qualification of Alloy 709 as a Class A construction material in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors. This initiative includes a comprehensive Alloy 709 code qualification plan aimed at generating extensive material testing data crucial for compiling the code case data package. The data package is essential in establishing material-specific design parameters for Alloy 709 to be used as Section III, Division 5 Class A construction material for fast reactors, molten salt reactors and gas-cooled reactors. An ASME Section III, Division 5 material code case requires the evaluation of mechanical properties from a minimum of three commercial heats, covering anticipated compositional ranges. A key part of the data package involves fatigue and creep-fatigue testing at elevated temperatures, needed for developing the fatigue design curves and the damage envelope of the creep-fatigue interaction diagram (D-diagram). This paper summarizes the strain-controlled fatigue testing on three commercial heats of Alloy 709 at 760 and 816°C with strain ranges between 0.25% and 3%. The fatigue failure data are used to generate a preliminary fatigue design curve. Additionally, the creep-fatigue testing results at 816°C with tensile hold times of 10, 30, and 60 minutes are presented in support of developing the D-diagram for Alloy 709.

Supercritical carbon dioxide cooling during machining has been identified as an effective measure to mitigate the risk of stress corrosion cracking in materials utilized in the primary circuit of light water reactors, particularly in pressure vessel structural steels. This study aims to compare two different cooling methods, the novel supercritical carbon dioxide and conventional high pressure soluble oil, employed during both milling and turning processes for SA508 Grade 3 Class 2 and AISI 316L steels. As the surface conditions of materials are critical to fatigue properties, such as crack initiation and endurance life, the fatigue performance of both cooling methods for each process were then evaluated and the impact on properties determined. To compare the potential benefits of supercritical carbon dioxide cooling against conventional soluble oil cooled machining, fatigue specimens were machined using industry relevant CNC machine tools. Surface finish and machining methods were standardized to produce two different specimen types, possessing dog- bone (milled) and cylindrical (turned) geometries. Force-controlled constant amplitude axial fatigue testing at various stress amplitudes was undertaken on both specimen types in an air environment and at room temperature using a stress ratio of 0.1. The fatigue performance of the supercritical carbon dioxide cooled specimens revealed substantially greater endurance lives for both SA508 and 316L materials, when compared with specimens machined using high pressure soluble oil cooling.

A compositional modification has been proposed to validate an alloy design which potentially eliminates the requirement of post-weld heat treatment (PWHT) while preserving the advantage of mechanical properties in a reduced activation bainitic ferritic steel based on Fe-3Cr-3W-0.2V- 0.1Ta-Mn-Si-C, in weight percent, developed at Oak Ridge National Laboratory in 2007. The alloy design includes reducing the hardness in the as-welded condition for improving toughness, while increasing the hardenability for preserving the high-temperature mechanical performance such as creep-rupture resistance in the original steel. To achieve such a design, a composition range with a reduced C content combining with an increased Mn content has been proposed and investigated. Newly proposed “modified” steel successfully achieved an improved impact toughness in the as- welded condition, while the creep-rupture performance across the weldments without PWHT demonstrated ~50% improvement of the creep strength compared to that of the original steel weldment after PWHT. The obtained results strongly support the validity of the proposed alloy design.

A thorough understanding of interactions between graphite and fluoride fuel salts is crucial, as graphite is a promising candidate for the moderator of molten salt reactors. This study investigates the infiltration of fluoride fuel salts into graphite and the fluorination of graphite by these salts under various pressures and temperatures. A high-pressure salt infiltration test apparatus was developed to examine the infiltration of NaF-KF-UF 4 and NaF-BeF 2 -UF 4 -ZrF 4 fuel salts into two types of graphite at high temperatures. For tests using NaF-BeF 2 -UF 4 -ZrF 4 , two different temperatures were selected to assess the impact of temperature on threshold pressure. The study observed salt infiltration into graphite at pressures exceeding its threshold pressure, and the threshold pressure for infiltration was lower at the higher temperature. In addition, the formation of carbon fluorides on the surface of post-test graphite specimens was identified.