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Molten Salt Reactors are described and more detail is given about actinide molten salts in particular, as well as molten chloride salts. Instead of micro-structure evolution, the challenge for MSRs is chemical evolution due to fission and corrosion products and their impact on chemistry, corrosion, phase equilibria, and thermophysical properties.
Development of non-power molten salt reactors (MSRs) are under consideration to further establish an MSR experience base, support the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.43(e), and provide any additional analyses needed for development of a full-scale MSR. Guidance provided in this report is based on MSRs operating with liquid fuel (i.e., fuel dissolved within a molten salt). These reactors, unless owned by the DOE or DOD, will require licensing by the US Nuclear Regulatory Commission (NRC) staff. Standard review plan (SRP) guidance for large light water reactors (LWRs) is available in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants; Light Water Reactor (LWR) Edition. However, NRC staff observed that NUREG-0800 is very cumbersome to apply to non-power reactors “because of the great differences in complexity and hazards between non-power reactors and nuclear power plants.” Therefore, a program to develop performance-based guidance applicable to non-power reactors was initiated. In 1996, NUREG-1537, Parts 1 and 2, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, was published. Part 1, the format and content guide, suggests a uniform format for presenting information in non-power reactor applications that is acceptable to the NRC staff, but conformance with the format and content is not required. Part 2, the SRP, ensures the quality and uniformity of the staff review of an application. Unfortunately, the application guidelines and SRP do not provide adequate guidance for all advanced non-LWR technologies and applications. This discrepancy eventually led to the 2012 development of interim staff guidance (ISG) for NUREG-1537, which includes criteria for describing and reviewing aqueous homogeneous reactors (AHRs). Specifically, NUREG-1537 ISG, 2012 expanded the original document to address three areas: 1. updated criteria for heterogeneous non-power reactors, 2. criteria for licensing AHRs, and 3. criteria for licensing a Part 50-licensed isotope production facility. In 2015, the US Department of Energy (DOE) opted to build on the AHR NUREG-1537 ISG experience by performing a gap analysis of the guidance that would be used to license a non-power MSR. MSRs represent one of the advanced non-LWR technologies selected by DOE for development through a multiyear cost share award with Southern Company Services. Under this Advanced Reactor Concepts 2015 (DOE Advanced Reactor Concepts [ARC] 15) award program, the DOE tasked Oak Ridge National Laboratory (ORNL) to evaluate the guidance changes that the NRC may need to consider when licensing an MSR non-power reactor. ORNL staff, with support from Boston Government Services, LLC, focused on five system-related chapters in NUREG-1537 that were considered most relevant to inform the effort that would be required for a non-power MSR applicant. ORNL documented this review in a technical report, ORNL/TM-2018/834, Proposed Guidance for Preparing and Reviewing Molten Salt Non-Power Reactor License Applications (NUREG-1537). The report was subsequently shared with industry and the NRC. The 2018 review was limited in scope, focusing on key system chapters based on the expected significance of each chapter relative to expected differences in addressing advanced non-LWR technologies, specifically non-power MSRs, compared with heterogeneously fueled non-power reactors. In the ORNL report, proposed generic adaptations were suggested for the following NUREG-1537 chapters: Chapter 4, “Reactor Description”; Chapter 5, “Reactor Coolant Systems”; Chapter 6, “Engineered Safety Features”; Chapter 9, “Auxiliary Systems”; Chapter 11, “Radiation Protection Program and Waste Management” The inclusion of Chapter 11 in the previous review effort was intended to provide guidance for categorizing the waste-handling process for an MSR operating with homogenous fuel. The introductions from Parts 1 and 2 of the 2012 NUREG-1537 ISG provide guidance for the application and review of production facilities. After a period of operation, non-power MSRs with homogenous fuel will include gaseous and soluble fission products. The gaseous fission products will be collected and held for decay in an off-gas system. There might also be an initiative to polish or filter the soluble fission products in the fuel salt by some mechanical or chemical means. The treatment and handling of fission products in the non-power MSR fuel salt and the description of this process in the safety analysis report (SAR) must be very precise to avoid the waste treatment facility being construed as a co-located special nuclear material (SNM) fuel cycle facility (see Section 2.3 of this report). Subsequent to the release of ORNL/TM-2018/834, NRC staff expressed a desire to continue the regulatory gap analysis that was begun in that report. This would provide additional clarity and information addressed in certain sections of the original report, while also providing new guidance on certain topics not addressed in the original report. This revision would benefit the NRC staff reviewing applications involving non-power MSR designs and would help developers understand how the NRC staff might approach the review of such applications. The focus of this report is to provide infrastructure support to the NRC staff for the regulatory review of non-power MSRs.
Thermal properties of molten salts, both with and without fissile and fertile material, are fundamental to modeling and simulation that supports design, construction, operation, and accident progression evaluation of molten salt reactors. This roadmap provides guidance to the research community on the types and amounts of required data for base, or carrier, salt systems, recommends simulated burnup fuel salt compositions to evaluate, and provides a technical foundation for a focused salt property evaluation matrix. The property values generated from both computation and experimental methods are used to develop models that constitute the Molten Salt Thermal Properties Database (MSTDB). While the building blocks of MSTDB are fundamental in nature, its data facilitates constructing engineering models that allow stakeholders to investigate the behavior of specific compositions of interest and supports broader modeling and simulation of MSRs through coupling to multi-physics, multi-scale mass accountancy tools.
Molten salt reactors (MSRs) are an attractive advanced reactor technology because of their increased fuel utilization and efficiency, the ease of fueling, and the potential to use alternative fuel cycles, including thorium and breeder/burner scenarios. Since an MSR has never been domestically operated for commercial electricity production, minimal infrastructure exists to support performance and materials testing, reactor licensing (including research, test, and operating reactors), and technology development in general. These factors could present challenges for the commercialization of the proposed design technology. Several research and development efforts and investments have begun to identify and supplement the infrastructure to accommodate the molten salt concept, including the United States Department of Energy Office of Nuclear Energy Molten Salt Reactor Program. This report reviews safeguards technology and domestic infrastructure to support the evaluation of nuclear safeguards (primarily material control and accounting) measures and approaches for salt-fueled MSRs. Recommendations for potential priority infrastructure investments, including safeguards measurement test beds and future research, are presented.
To support the development and utilization of the SAM code for molten salt-cooled pebble bed reactor safety analyses and licensing, selected improvements were made to the one-dimensional system-level heat transfer modeling in SAM with the goal of better capturing heat transfer phenomena specific to these designs. In consideration of molten salt as a coolant, new models were introduced to capture the enhanced radiative heat transfer that must be considered at the higher working temperatures. Radiative heat transfer is handled not only between two structural surfaces, but also in combination between the surfaces and the molten salt fluid, which has a higher opacity and absorption of radiative heat as compared to water. Several examples of analytical verification of the new radiative heat transfer methods are included in this report. Additionally, convective heat transfer correlations were selected from literature and included in SAM to account for the enhanced heat transfer expected in pebble bed geometry flows. These correlations are verified and demonstrated in the present work with a reference FHR example.
Perovskite phases have been explored in recent literature as waste forms for long term storage of radioactive waste from molten salt reactors specifically. Synthesis of new halide perovskite phases can inform these efforts as well as molten salt reactor chemistry in general. Inorganic perovskites have also shown promise as new scintillating materials for detection of X-rays and gamma rays. Modifying the structure and chemistry of these compounds, perhaps with organic components, could be used to synthesize new scintillating compounds or modify the properties of existing ones. Both U 6+ and U 4+ readily form compounds with a striking similarity to vacancy ordered perovskites (Cs 2 UO 2 Cl 4 and Cs 2 UCl 6 ). Each of these compounds is relevant to molten salt reactor chemistry and/or waste streams. These phases could from in uranium chloride salt systems or waste streams: Cs being a fission product.
An emerging new class of promising materials, High-Entropy Ceramics (HECs) or Compositionally-Complex Ceramics (CCCs), have been successfully fabricated in the last three years. This project evaluates HECs/CCCs in the concentrated solar power (CSP) context. In this project, we have identified a couple of simple, and key descriptors governing thermal conductivity in HECs. Size disorder among the cations was found to have a strong influence on the thermal conductivity across multiple systems and throughout a vast compositional space. Since identifying this, we have optimized previous compositions to achieve low thermal conductivities of $k$~$1-1.4Wm{-1}K{-1}$. We have further reduced our selection of promising compositions based on project guidelines and applied a theory that can potentially predict reactivity with molten salts to identify a small set of HECs that are suitable for further molten salt heat transfer media (HTM) stability testing. Many compositions reveal poor resistance to molten chloride salt after being exposed for 24 h compared to the reference structural alloy, Haynes 230. However, after 100 h, a few HECs/CCCs show strong parabolic rate constant effects and provide greater resistance to chloride salt compared to Haynes 230 alloy, which shows more linear kinetics. Postmortem characterization of the microstructure reveals the parabolic rate constant originates from a dense reaction layer form at the interface. This feature is absent in Haynes 230, which only shows grain boundary infiltration. The HECs/CCCs were also put in contact with the metal alloy under Ar for 800 h at 800°C and show negligible interaction. Lastly, parabolic rate constant HECs/CCCs also have low thermal conductivity $(<2 Wm^{-1} K^{-1})$ at 800°C. These specimens satisfy all project requirements and demonstrate their promising application in next-generation CSP.
This report describes the technologies that were developed as part of the Continuous Particle Monitoring and Removal for Molten Chloride CSP Systems project that was funded by the EERE Solar Energy Technologies Office during FY2021. This project consisted of two main tasks: (1) the development electrical resistance tomography (ERT) sensors to provide particle sensing capabilities for CSP-relevant molten chloride salts, and (2) the development of vortex separation technologies to enable efficient removal of particles from the salt. Both of these tasks are crucial because, although the Gen3 CSP chloride heat transfer fluid is inexpensive and can achieve high temperatures, major issues with erosion corrosion and clogging can arise from the presence of MgO particles within the salt. The particle monitoring and removal technologies were both demonstrated over the course of this project. The ERT sensors were able to achieve their performance targets by providing accurate particle concentration measurements over a range from 0.0 to 8.0 wt%. The vortex separators were able to demonstrate separation of particles from molten salt streams, but the separation efficiencies were lower than the targeted level of 90%. With further development and deployment, these technologies will ultimately help to enable successful long-term operations of CSP systems that make use of high-temperature chloride salts.
In assessing molten salt compatibility with structural materials, static testing is often reported because it is easy to conduct. However, in order to assess mass transfer, where material dissolves in the salt in the hot section and may precipitate out of the liquid in colder regions, a flowing salt experiment in a temperature gradient is needed. While a forced convection or pumped loop produces prototypic flowing conditions, it also is a multi-million dollar investment. Among the less expensive dynamic testing options, the thermal convection loop (TCL) was favored during the 1960’s development of molten salt reactor technology. Because the workforce and core facilities are currently available to conduct TCL experiments, it is the least expensive strategy to evaluate candidate structural alloys in flowing molten salt conditions.
As presented in this report, thermal conductivity and viscosity measurements were performed on key chloride pseudo-binary molten salt systems of relevance to molten salt reactor developers. Thermal conductivity measurements were conducted with a variable gap technique, in which a known heat flux is driven across a molten salt specimen and the temperature difference is measured, allowing calculation of the thermal conductivity. This is achieved by establishing a small gap between the bottom of a cylindrical inner containment, which houses electrical heating elements, and an outer containment, which houses cooling channels; the gap size can be varied by compression of a formed bellows. A new calibration scheme was developed herein, involving a correction factor to the heat flux based on He measurements at various temperatures. Furthermore, the data processing methodology was improved to minimize the impact of radiative heat transfer in the calculation of salt specimen thermal conductivity from the temperature difference measurements. Viscosity measurements were conducted with a rolling ball viscometer, in which a ball rolls some known distance in an angled tubular crucible, and the terminal velocity can be used to calculate the viscosity of the salt. The measurement can be performed in a quartz crucible, with which a standard camera can be used to track the ball, or in a metal crucible, with which x-ray radiography is required to track the ball. A new custom x-ray system was made and dedicated to the rolling ball viscometer to enable high throughput automated measurements with salts which require containment with metal. Both systems have been integrated with a ventilation stack which allows for off-gassing of radioactive material, enabling future measurements with U-bearing salts. The thermal conductivity measurements performed herein were with NaCl-KCl (44 mol% NaCl). This salt system was measured in the previous fiscal year, however the thermal conductivity values obtained were comparatively low, and so the system was remeasured with the aforementioned calibration scheme and improved post-processing techniques. The newly obtained thermal conductivity values for NaCl-KCl (44 mol% NaCl) indicate good agreement with kinetic theory and ab-initio models (within 5–10 %). The viscosity measurements performed herein were with three different compositions of NaCl-KCl: 75, 50, and 25 mol% NaCl; a new calibration scheme was applied to account for variable flow effects in the laminar regime. The results show reasonable agreement with literature (5–20 %, depending on the temperature, composition, and study); however, literature values are likely higher than true values based on pure end-member measurements performed in the comparative studies. The measurements conducted herein do show a trend such that viscosity increases with increasing NaCl concentration, which agrees well with one of two comparative studies. In general, the measurements conducted herein gives confidence in the capability to use these systems to accurately measure thermal conductivity and viscosity of actinide-bearing salts within the next fiscal year.
The Department of Energy (DOE)’s NEAMS focuses its efforts on the development of advanced modeling and simulation (M&S) tools for light-water reactors (LWRs) and non–LWRs (i.e., molten salt reactors, high-temperature gas reactors, microreactors, and fast reactors). In the previous fiscal year, the Multiphysics Applications Driver Technical Area funded molten salt reactor (MSR) M&S at Oak Ridge National Laboratory (ORNL) to generate multigroup macroscopic cross sections with Shift for a MSRE 2D lattice model in Griffin. In addition, Shift’s capability to calculate gamma dose rates from activated components in the primary exchangers in a molten salt breeder reactor was also demonstrated. In fiscal year 2023, ORNL generated multigroup macroscopic cross sections using Shift for a 3D MSRE core model. MSRE depletion calculations using Griffin were also demonstrated in this fiscal year. For the depletion calculation, one-group microscopic cross sections for the 3D MSRE core were generated using Shift, and the decay transmutation library from ORIGEN was converted to an ISOXML file, which is required as input in Griffin. Several Monte Carlo codes, such as OpenMC and Serpent, were also used to benchmark and supplement multigroup cross sections generated by Shift. Multigroup libraries were generated with 8 and 20 group structures, and the study found the 8-group structure to be more accurate when comparing Griffin results to continuous energy (CE) Monte Carlo results. The average flux from CE Shift calculations is up to ~6% higher than the CE Serpent calculations because of different values applied for the energy released per fission (κ values). The average flux in the fuel salt calculated by Griffin using cross sections generated with Shift agrees well with the reference CE Shift solution; the same is valid for the corresponding Serpent results. The maximum relative error is ~6% and ~2% compared to the CE Shift and Serpent reference solutions, respectively. Meanwhile, the average flux calculated by Griffin in the graphite moderator shows a higher difference in the thermal range when compared to both reference Monte Carlo solutions; this result suggests a need for improvement in cross section generation for the graphite moderator in the thermal range in both Monte Carlo codes. Griffin depletion calculations using cross sections from Shift and Serpent were performed and compared against ORIGEN calculations, and the nuclide densities calculated by Griffin were found to be generally in agreement with those of ORIGEN. Because a different approach was taken to calculate the energy released per fission ( κ values) in Shift and Serpent, a difference in nuclide densities from differences in the average flux was observed between Griffin using Serpent and Shift cross sections. Griffin calculations with Shift cross sections produced higher average flux in the salt than with Serpent cross sections, leading to higher consumption of 235 U and higher production of 135 Xe. For time-dependent depletion calculations, cross sections were generated with Serpent, and Griffin’s results using these cross sections were compared to CE Serpent depletion results, demonstrating good agreement. The average difference in keff between Serpent and Griffin as a function of burnup is about 155 pcm. Similarly, good agreement with small differences up to ~0.5% was also noticed in the nuclide density of 235 U and 135 Xe. More details regarding the methodologies invoked to generate the cross section to make code-to-code comparisons are discussed further in this report. User feedback on Griffin and Shift capabilities that will enhance these calculations is provided in this report for future consideration. The work performed this fiscal year can be extended further for multiphysics coupling of Griffin-Pronghorn/SAM with Mole to study precursor flow and salt chemistry.
Knowing thermophysical and thermochemical property values of salts over the expected range of operating temperatures is essential to modeling and simulation efforts that support the commercialization of molten salt reactor (MSR) technologies. Salt properties being measured at Argonne include thermal transitions, phase behavior, heat capacity, density, surface tension, volumetric thermal expansion, thermal diffusivity, thermal conductivity, and viscosity. The properties of several salts of interest including NaCl-UCl 3 and LiF-NaF-KF (FLiNaK) have been measured at Argonne and reported previously. These property measurements are suitable for use in evaluating reactor performance during startup and the early operating life of the reactor. The physical and chemical behavior of the fuel salt is altered by the buildup of fission products and the resulting changes in heat transfer and other properties must be understood by reactor developers. This report summarizes melting behavior, heat capacity, and thermal diffusivity values that were measured for mixtures of FLiNaK and NaCl-UCl 3 salts that have been doped with surrogate fission products. Additional heat capacity measurements were also performed on a separate mixture of FLiNaK without dopants. Measured properties are suitable for incorporation into the Molten Salt Thermal Properties Database–Thermochemical (MSTDB–TC) and the Molten Salt Thermal Properties Database–Thermophysical (MSTDB–TP) to support the development of MSRs. Salts were prepared by doping eutectic mixtures of NaCl-UCl 3 and FLiNaK with surrogate fission products. These salts were prepared for use in salt spill experiments being performed at Argonne and were subsequently used for thermal property measurements. Thermal transitions and heat capacities were measured by using differential scanning calorimetry (DSC) and thermal diffusivity was measured by using laser flash analysis (LFA) at temperatures spanning the typical MSR operating range. The measured property values of the doped salts were compared to measured property values of salts without fission products to quantify the effects of the doped fission products.
This report details an assessment of the hazards and safety impacts related to front-end operations of molten salt reactors (MSR) as well as an assessment of existing safety guidance. Publicly available information was evaluated to make recommendations to the U.S. Nuclear Regulatory Commission (NRC) for guidance development and information gathering needs. Five different MSR technologies were evaluated along with information from 1960’s about the Atomic Energy Commission programs for the Aircraft Reactor Experiment and Molten Salt Reactor Experiment. Primary areas of evaluation include fuel cycle facilities, transportation and at reactor fuel mixing operations up to reactor startup. In general, there was difficulty synthesizing publicly available information due to the early stage of the MSR development and proprietary information associated with the reactor technology companies analyzed. There are a few specific recommendations from this report. One overarching paradigm shift that the NRC needs to evaluate is the possibility of collocating fuel cycle facilities with operating MSRs. Although there is no specific prohibition to this in the NRC regulations, the collocation of the fuel cycle facilities with the operating MSR presents a novel approach compared to the light water reactor baseline. More guidance and research into thorium operations and regulation is needed. The hazards associated with the stability of stored fuel constituents need to be evaluated. Some reactor fuel cycles use beryllium, and its use and the associated hazards need further analysis. NRC-approved designs for transportation packages for shipping molten salt fuels and factory-built and fueled reactors are not available.
The Liquid Salt Test Loop (LSTL) at the US Department of Energy’s (DOE’s) Oak Ridge National Laboratory (ORNL) was developed to demonstrate technology for high-temperature fluoride salt systems (Figure 1). The LSTL is primarily constructed using Inconel 600 alloy and operates at temperatures of up to 700°C. The facility is loaded with 165 kg of LiF-NaF-KF salt (FLiNaK). This salt provides a relevant test environment for de-risking technology while avoiding the costs and hazards associated with beryllium-based or uranium-bearing salts. FLiNaK is also an advantageous salt for the secondary side of molten salt reactors. The facility’s major components include a centrifugal pump for salt circulation, an air-based heat exchanger to reject heat, a suite of instrumentation, and trace heating to prevent salt freezing. Additional heating is available through an induction heater rated at 200 kW. The relatively large heating and cooling capability enables the formation of a temperature gradient across the loop (i.e., a hot and a cold side), which is important for chemistry and corrosion studies. The LSTL is a unique US capability for high-temperature molten halide salt testing. Although some efforts are underway at universities, the LSTL’s scale, co-located purification system, and relatively large power differentiates it from other testing systems. Furthermore, unlike efforts within industry, access to the DOE-supported facility and communication of results, which are generally disseminated publicly, result in a broad significance in the molten salt reactor community.
Molten salt reactors (MSRs) are a class of Generation IV advanced reactor technologies aimed to enhance and improve the safety, fuel utilization and cost-effectiveness of nuclear power generation. MSR concepts are based on using a molten salt mixture as a primary nuclear reactor coolant, while the fuel can be either directly dissolved in the coolant (i.e., fluid-fueled) or can be in a separate solid form (i.e., salt-cooled solid-fueled). MSRs can be operated in either thermal or fast neutron spectra, as simple fissile convertors, or breeder reactors (e.g., utilizing both fissile and fertile fuels), as well as accelerator-driven sub-critical reactors or transuranic element burners. Operation of MSRs will necessitate management and treatment of the off-gas streams released to the headspace of the reactor where, in some designs, a cover gas can be circulated to remove certain fission products and maintain an inert atmosphere. An inert MSR cover gas swept over or sparged into the salt will confine radionuclides emerging from the free surface, including radioactive noble and non-noble gases, aerosols, and volatile species. The elements and compounds of these species will have a wide range of thermochemical and physical properties, some of which will decay to different products in the off-gas stream. Therefore, system components used to manage and treat MSR off-gas streams will need to be robust under dynamic reactor and chemistry conditions to support adequate reactor performance by managing the composition and species in the salt coolant and cover gas space. Off-gas system components would perform as a radionuclide boundary to the reactor vessel and a primary means for preventing radionuclide release; therefore, they are of high safety significance for MSR operations. This report reviews various technologies and materials for potential implementation in the design of an MSR off-gas system per a general conceptual framework. The reviewed technologies operate based on scrubbing, capture, delay, or separation of species in the off-gas stream. These include molten hydroxide scrubbers, solid sorbents, delay off-gas systems, particle traps, and cryogenic distillation. Since aspects of an off-gas system will be design-specific, the approach is to provide general discussions on attributes of these technologies and materials, including target off-gas stream species, operational parameters and conditions of importance to decontamination factors, implementation maturity, system component monitoring, potential off-normal conditions and considerations for waste products. The information aims to support the U.S. Nuclear Regulatory Commission in their safety evaluations for MSR designs.
The safety and reliability of next-gen reactors, i.e. molten salt reactors (MSR), depends on materials performance and longevity in extreme environments, yet no in-situ, validated measurement techniques for environmentally assisted cracking (EAC), one potential degradation mechanism, exist for these conditions. This project combined the use of in-situ direct current potential drop (DCPD) crack growth rate determination applied to a horizontal load frame with high temperature fittings and novel sample geometries to enable evaluation in MSR environments and develop a full in-situ SCC measurement capability for extreme environments (high conductivity and elevated T), necessary to predict materials reliability. A successful demonstration of DCPD measurement in molten NaNO 3 /KNO 3 salts at 340 o C was accomplished and validated through post-test fractography imaging. This is the first ever public example, to the authors knowledge, of an in-situ crack growth rate measurement in a molten salt environment.
It is known that the predictive life of the refractory ceramic liner of nuclear waste glass melters is conservative, as demonstrated by performance of these materials such as in the Defense Waste Processing Facility (DWPF). The motivation for this task is to maximize the useful life of the melters that will be operated at the Waste Treatment and Immobilization Plant (WTP), which will in turn minimize procurement and disposal costs and melter outage times, as well as to identify maximum loadings in the waste glass of those species that corrode melter components. This task was initiated jointly with Pacific Northwest National Laboratory (PNNL) with the objective to develop a methodology and model to enable more accurate prediction of refractory service life under prototypic conditions from laboratory-scale material corrosion tests. Refractory corrosion is generally reported as physical material loss, measured in units of distance (e.g., inch) or as physical material loss rate, measured in units of distance per time (e.g., inch/day). In post-operational melters, the refractory corrosion is measured directly, sometimes reported as corrosion depth. Crucible tests are used in the laboratory to accelerate the refractory corrosion to facilitate a meaningful measurement in a commensurate amount of time. Crucible tests are particularly useful in understanding refractory corrosion across a large glass composition space, where operational testing would be prohibitive. Some of the critical parameters that are known to influence refractory corrosion by molten glass in a crucible test are temperature, system redox, molten salt phases, glass chemistry, and test duration. The majority of data collected for Monofrax® K-3 (hereafter referred to as K-3) corrosion is from crucible tests, but a small amount comes directly from scaled and production melters. Crucible test data has been collected under varying conditions, whereas data collected from operational melters is relatively fewer and represents conditions specific to the melter campaign. The result is that the published data can be grouped and analyzed in multiple ways, not all of which are readily comparable. The Standard Test Method for Isothermal Corrosion Resistance of Refractories to Molten Glass (ASTM C621) outlines the general guidelines used across industry. That method describes a sealed, static test in which the surface area of the refractory coupon and the volume of glass are fixed. A significant portion of the crucible data pertaining to nuclear waste glasses has been collected in a modified configuration; the most notable differences being the surface area of the refractory coupon to volume of the glass and use of a method for bubbling the melt. To our knowledge, the influence of those parameters on the refractory corrosion has not been quantified. In this work, it was determined that static tests and bubbled tests would be performed. Savannah River National Laboratory (SRNL) was tasked with setting up and performing static testing while PNNL was tasked with setting up and performing bubbled testing. Initial activities were performed to establish laboratory methods that reproduce data comparable to existing data sets of K-3 refractory corrosion by low activity waste (LAW) and high-level waste (HLW) glass compositions. Later activities were focused on refining the test parameters to establish a standard test practice to be used between Laboratories and collecting additional data to be used in the enhanced waste glass model development. This document serves primarily to convey the refractory loss measurement results from corrosion testing of K-3 refractory with waste glass compositions developed for use in the WTP melters. The data will be used in the enhanced property/composition models being developed for waste glass vitrification and melter operations.
The demand for higher system thermal efficiencies requires the operation of power generation cycles and heat conversion systems at progressively higher temperatures. As the system operating temperature increases, existing materials may not provide adequate mechanical properties or environmental compatibility or both. There is an increasing commercial interest in the development and deployment of liquid-fueled Molten Salt Reactors (MSRs). Hastelloy®N, the highest performing candidate MSR structural alloy, is not capable of operations at temperatures above 700°C, thus limiting the performance of these systems. Using an Integrated Computational Materials Engineering (ICME)-approach and small laboratory scale heats, ORNL developed a class of patented alloys covered by U.S. Patent 9,435,011 B2, “Creep-resistant, Cobalt-free alloys for high temperature, liquid-salt heat exchanger systems,” similar to Hastelloy®N in that they are primarily solid solution strengthened. In contrast to precipitation strengthened alloys, the microstructure of solid solution alloys and hence the high temperature mechanical properties are stable for extended periods of time allowing long reactor operating life. The new alloys have shown to possess good resistance to liquid fluorides at temperatures up to 850°C and have significantly improved creep properties when compared to Hastelloy®N. The purpose of the CRADA project was for ORNL to collaborate with Haynes International- a materials producer, MetalTek International- a foundry, and Kairos Power – an advanced reactor developer – to scale-up selected alloys, evaluate their properties, and identify one solid solution strengthened alloy that can meet the property requirements for the reactor being developed by Kairos Power and other similar liquid fluoride-salt cooled reactors. As part of the project, eight alloys were down-selected and fabricated in larger industrial scale heats by Haynes International. Resistance to molten salt was evaluated in flowing FLiNaK and FLiBe by Kairos Power using their Rotating Cage Loop (RCL) system. Accounting for iron deposition during these tests, the new alloys displayed very low net mass change showing excellent corrosion performance in molten salt. Creep properties evaluated at ORNL were found to be better than that of Hastelloy®N and 316 stainless steel. Long-term stabilities of the alloys evaluated by Haynes International showed that these alloys have excellent thermal stability in the temperature range 704.4-815.6°C, with the change in strength and ductility being less than 10-15% after a 4000-hour exposure at 815.6°C. Autogenously Gas Tungsten Arc Welding (GTAW) welded samples showed less than 10% change in yield strength / ultimate tensile strength / total elongation compared to the basemetal, indicating that the alloys have excellent weldability. Three parts were successfully investment-cast using one alloy with very little voiding showing feasibility of fabricating parts using the casting process. This project enabled extensive interaction between the material producer Haynes International, casting supplier MetalTek, and reactor developer Kairos Power. This facilitated testing of materials and components produced using the newly developed alloys by the end-user. This allowed the generation of critical dataset required for down-selection of a few promising alloys for further development. This data is also currently being shared with other reactor designers for them to evaluate the suitability of this alloy for their reactor design. The availability of this alloy will ultimately enable the design and development and deployment of MSRs with increased temperature of operation and thus, improved efficiencies.