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.

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.

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.

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.

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.

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.

Diffusion bonded compact heat exchangers have exceptionally high heat transfer efficiency and might significantly improve the performance and reduce the cost of supercritical carbon-dioxide Brayton cycle power plants using high temperature heat sources, like high temperature nuclear reactors and concentrating solar power plants. While these heat exchangers have an excellent service history for lower temperature applications, considerable uncertainty remains on the performance of diffusion bonded material operating in the creep regime. This paper describes a microstructural modeling framework to explore the plausible mechanisms that may explain the reduced creep ductility and strength of diffusion bonded material, compared to wrought material. The crystal plasticity finite element method (CPFEM) is used to study factors affecting bond strength in polycrystals mimicking diffusion bonded microstructures. Additionally, the phase field method is also employed to simulate the grain growth and recrystallization at the bond line to model the bonding process and CPFEM is used to predict the resulting material performance to connect processing parameters to the expected creep life and ductility of the material, and to study potential means to improve the structural reliability of the material and the resulting components by optimizing the material processing parameters.

There is an increased interest in miniature testing to determine material properties. The small punch test is one miniaturized test method that has received much interest and is now being applied to support the design and life assessment of components. This paper presents the results of a test program for a small punch creep test at 650°C of 316L stainless steel produced from additive manufacturing. A major finding is that the deflection rate curve versus time may have multiple minima as opposed to forged 316L with only one minimum. This is believed to be due to microcracking and has direct consequences on the determination of the creep properties that that are based on a single minimum value in the CEN Small Punch Standard. In the paper, aged and nonaged materials are compared, and small punch creep results are also compared with standard uniaxial creep tests. The multiple minima feature means that the approach to determine equivalent stress and strain rate from the minimum deflection rate needs to be modified. Some approaches for this are discussed in the paper. Under the assumption that the multiple minima represent cracking, it opens up opportunities to quantify reduced creep ductility by the small punch test.

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.

As many nuclear power plants are in the license renewal operating period and some are entering subsequent license renewal, there is increased probability that repairs will be needed on components that have been exposed to significant neutron fluence. The neutron-driven transmutation of nickel and tramp boron in austenitic materials commonly used in reactor internals can lead to the generation of trapped helium and the associated risk of helium-induced cracking (HeIC) during weld repairs. In the weld heat affected zone, where temperatures are insufficient to allow the helium to diffuse out of the material, the helium can remain trapped. Upon cooling, the residual stresses, combined with weakened grain boundaries due to helium coalescence, can lead to cracking. The current ASME limit for helium content for Code repairs is 0.1 appm. Prior work has demonstrated a strong inverse correlation between helium content and permissible weld heat input for avoidance of HelC. The helium concentration in the material to be repaired is thus a critical input to the development of weld repair processes to be applied to these materials. The reliable measurement of helium in irradiated materials at concentrations relevant for the evaluation of HeIC risk is a specialized process. It is important to demonstrate that the capability is available and can be practically leveraged to support emergent repairs. This paper presents on the execution and results of a multi-laboratory test program aimed at demonstrating the industry capability of acquiring accurate, repeatable, and timely measurements of relatively low concentrations of helium (< ~20 appm) within austenitic materials commonly used in reactor internals. Participating laboratories were supplied with equivalent specimens extracted from boron-doped coupons that were irradiated to drive the boron-to-helium transmutation reaction. The results and lessons learned from the program are expected to support the development of industry guidance for the acquisition of similar measurements supporting nuclear component repairs.

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.

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.

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.

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.

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.

The Chinese power industry has experienced rapid development in the past decade. The newly built 600+°C ultra-super-critical (UCS) fossil fire power plants and pressed water reactor nuclear power plants in China are the world’s most advanced level technically and effectively. The available capacity of 600+°C UCS fossil fire power plant in China is more than 200 GW by the end of 2015, which has greatly contributed to the energy-saving and emission-reduction for China and the whole world. In China, the 610°C and 620°C advanced USC (A-USC) fossil fire power plants had been combined into the grid, 630°C A-USC fossil fire power plant is about to start to build, the feasibility of 650°C A-USC fossil fire power plant is under evaluation, 700°C AUSC fossil fire power plant has been included in the national energy development plan and the first Chinese 700°C A-USC testing rig had been put into operation in December 2015. The advanced heat resistant materials are the bottlenecking to develop A-USC fossil fire power plant worldwide. In this paper, the research and development of candidate heat resistant steels and alloys selected and/or used for 600+°C A-UCS fossil fire power plant in China is emphasized, including newly innovated G115 martensitic steel used for 630°C steam temperature, C-HRA-2 fully solid-solution strengthening nickel alloy used for 650°C steam temperature, C-HRA-3 solid-solution strengthening nickel alloy used for 680°C steam temperature, 984G iron-nickel alloy used for 680°C steam temperature, C-HRA-1 precipitation hardening nickel alloy and C700R1 solid-solution strengthening nickel alloy used for 700+°C steam temperature.

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.

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.

The Advanced Materials and Manufacturing Technologies (AMMT) program is aiming at the accelerated incorporation of new materials and manufacturing technologies into nuclear-related systems. Complex Ni-based components fabricated by laser powder bed fusion (LPBF) could enable operating temperatures at T > 700°C in aggressive environments such as molten salts or liquid metals. However, available mechanical properties data relevant to material qualification remains limited, in particular for Ni-based alloys routinely fabricated by LPBF such as IN718 (Ni- 19Cr-18Fe-5Nb-3Mo) and Haynes 282 (Ni-20Cr-10Co-8.5Mo-2.1Ti-1.5Al). Creep testing was conducted on LPBF 718 at 600°C and 650°C and on LPBF 282 at 750°C. finding that the creep strength of the two alloys was close to that of wrought counterparts. with lower ductility at rupture. Heat treatments were tailored to the LPBF-specific microstructure to achieve grain recrystallization and form strengthening γ' precipitates for LPBF 282 and γ' and γ" precipitates for LPBF 718. In-situ data generated during printing and ex-situ X-ray computed tomography (XCT) scans were used to correlate the creep properties of LPBF 282 to the material flaw distribution. In- situ data revealed that spatter particles are the potential causes for flaws formation in LPBF 282. with significant variation between rods based on their location on the build plate. XCT scans revealed the formation of a larger number of creep flaws after testing in the specimens with a higher initial flaw density. which led to a lower ductility for the specimen.

In this study, 25Cr2Ni2Mo1V filler metal was deposited to weld low pressure steam turbine shafts, which are operated in fossil power plants. A comparison experiment was conducted on the weld metals (WMs) before and after varied various aging duration from 200 hours up to 5000 hours at 350 ℃. Microstructure was characterized by means of scanning electron microscopy (SEM) and electron back-scattered diffraction (EBSD) techniques. In addition, mechanical properties of corresponding specimens were evaluated, e.g. Vickers microhardness, Charpy V impact toughness and tensile strength. It is shown that the tensile strength remained stable while impact energy value decreased with increasing aging duration. Based on the experiment above, it was concluded that the variation of mechanical properties can be attributed to the redissolution of carbides and reduction of bainite lath substructure.