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Overview of high temperature metals, including component construction rules for advanced reactor designs, assessment of filler metal for Alloy 800H weldments, high-temperature crack growth rate testing, work packages and contributors for FY24, creep-fatigue damage summation evaluation, GCR metal needs, and high temperature metals from other programs.
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Discussion of RELAP5-3D solutions to Exercise 1 of the OECD-NEA High Temperature Test Facility (HTTF) Benchmark. Results from ongoing benchmark activities supported by the Advanced Reactor Technologies - Gas-Cooled Reactor Program with use of the High Performance Computing Center at Idaho National Laboratory. Also discussion of Ultra Safe Nuclear Corporation and BWXT plans for the decade, and issues of modern thermal hydraulic tools for prismatic HTGR modeling. Open discussion of verification and validation and ART-GCR campaign to spearhead the development of an HTGR thermal hydraulics benchmark based on the HTTF.
We report SiC/SiC fiber composites are proposed cladding materials to improve the accident tolerance of commercial light water nuclear reactor fuel. To evaluate their viability, understanding the kinetics of corrosion under irradiation and high-temperature, high-pressure aqueous environment is critical. In the present work, the microstructure and mechanical properties of novel SiC fibers manufactured by Laser Chemical Vapor Deposition (LCVD) are introduced. The LCVD technology has control over microstructure and stoichiometry while allowing for high purity manufacturing, all key qualities needed in search of an optimum SiC fiber for nuclear energy applications. The isolated hydrothermal corrosion of commercially available carbon-rich Hi-Nicalon Type-S fibers is compared to the LCVD stoichiometric and LCVD silicon-rich fibers. Autoclave testing was carried out at 310°C and 14 MPa in the absence of irradiation. In a separate study, the effect of gamma irradiation on the fibers in the room-temperature environment was analyzed. It was observed from the experiments that the dissolution of the fibers was highly dependent on the stoichiometric ratio, fiber surface morphology and thermal pre-treatment of fibers. Contrary to previous findings, fiber dissolution was also visible under low-temperature gamma irradiation in aqueous environment with LCVD fibers showing lower dissolution.
ART Program First Quarter report for 2023. Entries for Fuels Development and Qualification, High Temperature Materials Development, Graphite Development and Qualification, and Design Methods Modeling and Validation.
This report summarizes various fuel types that may be applicable to several microreactor concepts, which are defined as (1) very high temperature reactors (VHTR), (2) sodium fast reactors (SFR), (3) system for nuclear auxiliary power (SNAP) reactors, (4) gas fast reactors (GFR), and (5) molten salt reactors (MSR). The fuel systems that were assessed include: uranium mononitride (UN), uranium monocarbide (UC), uranium dioxide (UO 2 ), uranium oxycarbide (UCO) Tristructural Isotropic (TRISO), UN TRISO, mixed oxide (MOX), metallic fuels, and metal hydrides. While UCO TRISO has undergone significant testing through the advanced gas reactor (AGR) program, the very high cost necessitates consideration of other fuel types. UN and UC were identified as fuels that should receive further investigation due to their thermophysical and mechanical properties. Minimal irradiation performance data shows that these fuels are good candidate fuels for microreactors.
Disposal of large, heat-generating waste packages containing the equivalent of 21 pressurized water reactor (PWR) assemblies or more is among the disposal concepts under investigation for a future repository for spent nuclear fuel (SNF) in the United States. Without a long (>200 years) surface storage period, disposal of 21-PWR or larger waste packages (especially if they contain high-burnup fuel) would result in in-drift and near-field temperatures considerably higher than considered in previous generic reference cases that assume either 4-PWR or 12-PWR waste packages (Jové Colón et al. 2014; Mariner et al. 2015; 2017). Sevougian et al. (2019c) identified high-temperature process understanding as a key research and development (R&D) area for the Spent Fuel and Waste Science and Technology (SFWST) Campaign. A two-day workshop in February 2020 brought together campaign scientists with expertise in geology, geochemistry, geomechanics, engineered barriers, waste forms, and corrosion processes to begin integrated development of a high-temperature reference case for disposal of SNF in a mined repository in a shale host rock. Building on the progress made in the workshop, the study team further explored the concepts and processes needed to form the basis for a high-temperature shale repository reference case. The results are described in this report and summarized..
Ferritic iron-chromium-aluminum (FeCrAl) alloys are an accident tolerant fuel candidate to replace the incumbent Zr-based claddings. Nuclear grade FeCrAl alloys are marked by superior high temperature mechanical behavior and exceptional steam oxidation resistance, both of which increase safety margins during accident scenarios. In the present study, the loss of coolant accident (LOCA) burst behaviors of three un-irradiated commercially fabricated cladding materials in a simulated LOCA environment were compared: (1) T35Y2, a 1 st generation nuclear grade FeCrAl, (2) C26M, a 2 nd generation nuclear grade FeCrAl, and (3) Zircaloy-2. Both FeCrAl alloys showed improved mechanical strength and steam oxidation resistance compared to Zircaloy-2. C26M claddings burst at significantly higher temperatures for all tested engineering hoop stresses, had limited ballooning, and demonstrated preferential fuel retention behavior in terms of burst opening area and length. High temperature tensile data for C26M is also presented. For both un-irradiated FeCrAl alloys, it was found that a distinct “threshold” burst stress signified the transition between small and large openings. Finally, higher threshold hoop stresses were associated with higher uniaxial strength for the FeCrAl alloys, indicating that tensile data, rather than creep data, could be useful for predicting rupture size and assessing fuel dispersal concerns.
There are in total 30NEUP projects focusing on the thermal-fluid experiments related with High-Temperature Gas-cooled Reactor (HTGR) from FY2009 to FY2021, producing a large amount of high-quality validation data, however, these valuable data has been scattering everywhere and not been disseminated to the community well, and final reports are available only from OSTI webpage. This could potentially be a huge waste, not only for government budget but also for the HTGR research community. To improve access to this HTGR validation data and optimize the return on the significant investment made by DOE and supported by the Advanced Reactor Technologies (ART) Gas-Cooled Reactor (GCR) program, we conducted a survey to assess completed and ongoing HTGR NEUP projects with the aim to develop a public-access data platform that can be used to retrieve code validation data and guide future NEUP investments.
Uranium (U) is often alloyed with molybdenum (Mo) or zirconium (Zr) in order to stabilize its high-temperature body-centered cubic phase for use in nuclear reactors. However, in all metallic fuel forms, the α phase of U remains in some fraction. This phase decomposition due to temperature or compositional variance can play an outsized role on fuel performance and microstructural evolution. Relatively little is known about fundamental point defect properties in α -U at non-zero temperatures, from either computational or experimental studies. This work performs the first thorough evaluation of the α phase of U via ab initio molecular dynamics (AIMD). A number of thermophysical properties are calculated as a function of temperature, including equilibrium lattice parameters, thermal expansion, and heat capacity. These results indicate a two-region behavior, with the transition at 400 K. The thermal expansion/contraction in the a / b direction occurs rapidly from 100 up to 400 K, after which a more linear and gradual change in the lattice constant takes place. The volumetric expansion matches experiments quantitatively, but the individual lattice constant expansion only matches experiments qualitatively. Point defect formation energies and induced lattice strains are also determined as a function of temperature, providing insight on defect populations and the fundamentals of irradiation growth in α -U. Interstitials induce significantly more strain than vacancies, and the nature of that strain is highly dependent on the individual lattice directions. The direction of point defect-induced lattice strain is contrary to the irradiation growth behavior of α -U. This work shows that isolated point defects cannot be the primary driving force responsible for the significant irradiation-induced growth of α -U observed experimentally.
The modeling of a loss-of-coolant-accident scenario involving nuclear fuels with FeCrAl cladding materials in consideration to replace a Zircaloy requires knowledge of the thermodynamics of oxidized structures. At temperatures higher than 1500 degrees C, oxidation of FeCrAl alloys forms (Fe,Cr,Al) 3 O 4 spinels. In situ high-energy X-ray diffraction in a conical nozzle levitator installed at beamline 6-ID-D of the APS was used to study the structural evolution of the oxides as a function of the temperature. Single-phase (spinel) and multiphase (spinel-corundum-FeAlO 3 ) regions are mapped as a function of the temperature for three different compositions of FeCrAl oxidation products. The thermal expansion coefficients and cation distribution in the spinel structure have been refined. The temperature at which complete melting of the fuel cladding is expected has been determined by the liquidus temperatures of the oxidized products to be between 1657 and 1834 degrees C in a 20% O 2 /Ar atmosphere using the cooling trace method. The liquidus temperature increases with increasing Al and Cr content in the spinel phase.
Small nuclear reactors, including small modular reactors (SMRs) or reactors for space applications, rely on different materials than those typically applied in large-scale nuclear power plants. Examples include molten salts as cooling medium or fuel carrier, metal hydrides as high-temperature moderators, and fuels allowing for higher burnup. All of these also require novel structural materials, for which material interactions have to be understood. Fissionable and fissile materials, such as uranium or plutonium, are rarely considered in materials design other than for nuclear fuels. Similarly, the aspects of radiation damage, occurring during irradiation when a reactor operates, are unique to nuclear materials research. The handling of these materials puts further limitations on the materials science conducted for nuclear materials. All of these issues move research for these materials off the “main stream” of materials science, and cause it to be more easily conducted at national laboratories. However, nuclear reactors offer unique opportunities for carbon-neutral energy generation and have great potential to address if not solve problems arising from global warming. This special topic, sponsored by the TMS Nuclear Materials Committee, focuses on materials research for small nuclear reactors, both experimental and simulation/modeling.
High-temperature microreactors can play a role in developing reliable, portable energy sources for off-grid remote locations, microgrid concepts, and industrial process heat. Portability and passive safety criteria tend to skew microreactor structural component designs toward complex geometries, high thermal stresses, and design bases with large numbers of startup/shutdown cycles. Current design rules, as typified by Section III of the American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel Code, are less than optimal for these conditions, particularly for preliminary component designs where developers need to rapidly consider a large number of potential component configurations. This paper presents a design method targeted toward rapid, efficient evaluation of preliminary component designs using modern finite element analysis. The new method retains key connections with the ASME Code rules and design data while streamlining the design approach. This paper presents the design method, several verification examples illustrating the similarities and differences between the new method and the current ASME rules, and the application of the new approach to the evaluation of a test article mimicking key features of a heat pipe–cooled microreactor.
Nuclear reactors have the potential to provide high energy density to enable sustainable surface power and advanced propulsion methods needed for human exploration activities at the moon and mars. Current mission planning is surveying different reactor types for space power and propulsion application. Of these reactor types, the use of a moderator within the reactor can enable reduced enrichment, reduce overall fuel loadings, and minimize the critical size of the reactor compared to unmoderated reference systems. This proceeding summarizes some moderator materials identified for space reactor applications: zirconium hydride, yttrium hydride, beryllium, and beryllium oxide, and the unique design considerations inherent to surface power and nuclear thermal propulsion reactor designs. It was found that there are four key considerations during the moderator selection and design process: nuclear properties, thermophysical & mechanical properties, manufacture & readiness, and environmental compatibility. Surface power reactors can benefit from moderators which minimize overall system mass and are capable of surviving high temperature irradiation environments for years with little degradation. Nuclear thermal propulsion reactors can benefit from moderators which are capable of retaining structural integrity under multiple burns while being exposed to a wide temperature range (40 < T < 500+ K). Moderator materials which exhibit good stability under irradiation and high temperature operation, minimize fuel pitch, and are high readiness are desirable for near term implementation.
Additive manufacturing (AM) methods are currently being explored for applications in nuclear reactors to make advanced reactors more efficient, safe, and reliable. The Transformational Challenge Reactor (TCR) program has explored AM for nuclear by designing a high-temperature gas reactor (HTGR) using an AM silicon carbide fuel form with uranium nitride–tristructural isotropic fuel. In this study, we detail the design process for the TCR fuel form’s coolant channels using computational fluid dynamics models with conjugate heat transfer. Additionally, this work discusses how these models were interfaced with other design teams, project milestones, and the agile design method used to mature the reactor design. The methodology deployed was able to create a channel design with lower maximum fuel temperatures and thermal stresses in the fuel form over traditional channel designs that can be manufactured subtractively. These results were achieved with only small manufacturing penalties. Results are discussed and presented on lessons learned for designing AM components for nuclear reactors. Finally, areas of opportunity are discussed for advanced design tools to further automate design activities and optimize reactors with fewer built-in assumptions.
The Microreactor Applications Research Validation and EvaLuation (MARVEL) project is producing a high temperature liquid metal-cooled nuclear test bed at Idaho National Laboratory (INL) to ultimately improve the integration of microreactors to end-user applications. This ambitious effort seeks to design, authorize, construct, test, and operate the reactor within five years. In order to construct and operate the MARVEL reactor in a timely manner, the system will utilize materials and component designs which have already been used, qualified, or licensed from previous reactors. The MARVEL reactor will be located at the INL Transient Reactor Test (TREAT) facility in the north high-bay equipment pit and will use the existing 304 stainless steel-clad U-ZrH1.6 pin-type fuel system developed by General Atomics and purchased from TRIGA International. This fuel has been previously qualified under the United States Department of Energy’s (US DOE) Reduced Enrichment for Research and Test Reactors (RERTR) Program. Even though the regulator of the MARVEL reactor is the US DOE, the standards and overall approach recommended by the Nuclear Regulatory Commission is well-defined and utilized here. Following NUREG-1537 regulatory guidance, this report documents the authorization case for the MARVEL fuel system’s application to MARVEL and establishes stable and predictable fuel performance during the most thermophysically unfavorable conditions achievable in the MARVEL reactor. To that end, this report provides a comprehensive survey of the known thermophysical properties, performance, and quantitative relationships associated with the MARVEL reactor fuel element and uses this information to determine its mechanical integrity and risk of reaching unacceptable conditions during the most extreme accident scenarios predicted for the reactor using the most conservative assumptions available. The information contained herein is compiled from a combination of historical reports and peer reviewed scientific publication manuscripts. Known mechanisms under which the fuel is susceptible to failure are highlighted and compared to conditions that could exist in the MARVEL reactor during an unanticipated transient or accident scenario. The two scenarios considered for analysis in this report are (1) an unprotected loss of flow accident and (2) a hypothetical unprotected loss of coolant accident during the loss of flow accident. Preliminary 2D steady-state analyses herein indicate that both fuel-cladding chemical interactions and fuel-cladding mechanical interactions are negligible throughout the fuel’s operational cycle under both normal and high temperature accident scenario conditions. Although higher fidelity 3D time-dependent modeling and simulations are planned, the following may be concluded presently. The MARVEL fuel element maintains its geometric stability and structural integrity during the most extreme accident scenarios predicted for the MARVEL reactor. The hoop stress during the unprotected loss of flow accident reaches about -1.27 MPa; this negative stress indicates that it is compressive rather than tensile. The compressive stress is a result of the NaK pressure on the outside of the fuel element, caused by the restricted thermal expansion of the NaK coolant, exceeding the internal pressures generated inside of the fuel element. The hoop stress generated in the cladding during the unprotected loss of coolant accident reaches a maximum of about approximately 10 MPa, which is nearly an order of magnitude less than the predicted yield strength of the cladding under high-temperature accident scenario conditions. Calculations were compared with results from high performance computational simulations using BISON and are in very close agreement. A conservative MARVEL fuel meat peak temperature limit of 900 °C is recommended presently, which is about 180 °C higher than the peak fuel temperature predicted to occur during the most extreme accident. Based on the known properties and behavior of the MARVEL fuel element, the fuel successfully meets its design and safety requirements under normal and most extreme accident conditions with a large safety margin.
The Microreactor Applications Research Validation and EvaLuation (MARVEL) project is producing a high temperature liquid metal-cooled nuclear test bed at Idaho National Laboratory (INL) to ultimately improve the integration of microreactors to end-user applications. This ambitious effort seeks to design, authorize, construct, test, and operate the reactor within five years. In order to construct and operate the MARVEL reactor in a timely manner, the system will utilize materials and component designs which have already been used, qualified, or licensed from previous reactors. The MARVEL reactor will be located at the INL Transient Reactor Test (TREAT) facility in the north high-bay equipment pit and will use the existing 304 stainless steel-clad U-ZrH1.6 pin-type fuel system developed by General Atomics and purchased from TRIGA International. This fuel has been previously qualified under the United States Department of Energy’s (US DOE) Reduced Enrichment for Research and Test Reactors (RERTR) Program. Even though the regulator of the MARVEL reactor is the US DOE, the standards and overall approach recommended by the Nuclear Regulatory Commission is well-defined and utilized here. Following NUREG-1537 regulatory guidance, this report documents the authorization case for the MARVEL fuel system’s application to MARVEL and establishes stable and predictable fuel performance during the most thermophysically unfavorable conditions achievable in the MARVEL reactor. To that end, this report provides a comprehensive survey of the known thermophysical properties, performance, and quantitative relationships associated with the MARVEL reactor fuel element and uses this information to determine its mechanical integrity and risk of reaching unacceptable conditions during the most extreme accident scenarios predicted for the reactor using the most conservative assumptions available. The information contained herein is compiled from a combination of historical reports and peer reviewed scientific publication manuscripts. Known mechanisms under which the fuel is susceptible to failure are highlighted and compared to conditions that could exist in the MARVEL reactor during an unanticipated transient or accident scenario. The two scenarios considered for analysis in this report are (1) an unprotected loss of flow accident and (2) a hypothetical unprotected loss of coolant accident during the loss of flow accident. Preliminary 2D steady-state analyses herein indicate that both fuel-cladding chemical interactions and fuel-cladding mechanical interactions are negligible throughout the fuel’s operational cycle under both normal and high temperature accident scenario conditions. Although higher fidelity 3D time-dependent modeling and simulations are planned, the following may be concluded presently. The MARVEL fuel element maintains its geometric stability and structural integrity during the most extreme accident scenarios predicted for the MARVEL reactor. The hoop stress during the unprotected loss of flow accident reaches about -1.27 MPa; this negative stress indicates that it is compressive rather than tensile. The compressive stress is a result of the NaK pressure on the outside of the fuel element, caused by the restricted thermal expansion of the NaK coolant, exceeding the internal pressures generated inside of the fuel element. The hoop stress generated in the cladding during the unprotected loss of coolant accident reaches a maximum of about approximately 10 MPa, which is nearly an order of magnitude less than the predicted yield strength of the cladding under high-temperature accident scenario conditions. Calculations were compared with results from high performance computational simulations using BISON and are in very close agreement. A conservative MARVEL fuel meat peak temperature limit of 900 °C is recommended presently, which is about 180 °C higher than the peak fuel temperature predicted to occur during the most extreme accident. Based on the known properties and behavior of the MARVEL fuel element, the fuel successfully meets its design and safety requirements under normal and most extreme accident conditions with a large safety margin.
Significant and major accomplishments of the Advanced Reactor Development: Gas-Cooled Reactor Program for the first quarter of 2021.