Results for “Cladding design” — Page 24

DOE OSTI2026

EDS Analysis of FCCI in AFC-FAST Fuel Pins

The Advanced Fuel Campaign’s Fission Accelerated Steady-state Testing (FAST) program uses metallic fuel pins with small diameters to reach a desired burnup more quickly. This enables accelerated testing of advanced fuel designs and decreases the time between idea conception and commercial usage. Post-irradiation examination is critical in this process, especially with respect to the fuel-cladding chemical interactions (FCCI). In this work, energy dispersive X-ray spectroscopy (EDS) is used to track how elements from the fuel and fission products have diffused through the cladding. Elemental redistribution along the fuel-cladding interface is mapped and FCCI region thicknesses are measured. The correlations between geometry, temperature, burnup, and FCCI thickness are presented.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

ATF Cladding Mechanical Properties Report: Capability Demonstration

This report documents mechanical testing capability demonstration activities performed in fiscal year (FY)2025 at Oak Ridge National Laboratory(ORNL)on chromium-coated (Cr-coated) and uncoated advanced zirconium alloy claddings irradiated in the High Flux Isotope Reactor (HFIR) to approximately 4 displacements per atom (dpa), corresponding to ~13GWd/t burnup. Specimens were prepared in axial tension (ATT) and ring tension (RTT) geometries, and passive silicon carbide(SiC)thermometry (TM) was employed to determine irradiation temperatures, which averaged 38–43 °C below the 330 °C design target. Mechanical testing at ambient temperature demonstrated the expected irradiation-induced hardening, with yield strength(YS)and ultimate tensile strength(UTS)values increasing substantially relative to unirradiated counterparts. However, this strengthening was accompanied by a reduction in ductility, as indicated by lower uniform and total elongations(UE and TE). Both coated and uncoated claddings exhibited similar mechanical response, though Cr-coated specimens showed surface cracking perpendicular to the loading direction, attributable to the hardness mismatch between the coating and substrate. Fracture in all cases remained ductile, and no coating spallation was observed following HFIR irradiation. Complementary efforts were directed toward the fabrication of test specimens from commercially irradiated cladding (rod 47I, ~31.1 GWd/t average burnup). Axial sectioning and computer numerical control(CNC)machining successfully produced ATT geometries suitable for benchmarking against HFIR-irradiated specimens. This capability enables direct comparison of cladding behavior between test reactor and commercial reactor environments, thereby supporting the validation of HFIR as a surrogate irradiation platform for accident tolerant fuel (ATF) development. Once HFIR irradiations are completed in FY26,the relevant comparison tests will be completed. Collectively, the FY 2025 PIE campaign has provided mechanical performance data for irradiated advanced claddings The demonstrated capabilities support the framework for mechanical testing and further evaluations in subsequent years. These efforts will represent an important contribution toward the licensing and deployment of Cr-coated zirconium alloy cladding as a near-term ATF solution.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Advanced Drying Cycle Simulator Thermal Response to Commercial Drying Conditions

The purpose of this report is to document updates on testing of the apparatus built to simulate commercial drying procedures for spent nuclear fuel at the Nuclear Energy Work Complex at Sandia National Laboratories. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system during subsequent storage and disposal. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates well-designed investigations of drying process efficacy and water retention that incorporate relevant physics and well-controlled boundary conditions. This report documents testing updates for the Advanced Drying Cycle Simulator (ADCS). This apparatus was built to simulate commercial drying procedures and quantify the amount of residual water remaining in a pressurized water reactor (PWR) fuel assembly after drying. The ADCS was constructed with a prototypic 17×17 PWR fuel skeleton and waterproof heater rods to simulate decay heat. These waterproof heaters are the next generation design to heater rods developed and tested at Sandia National Laboratories in FY20. This report describes preliminary testing of the ADCS through measurement and analysis of the thermal response of the system to a subset of commercial drying conditions that exclude the introduction of water, namely simulated decay heats and pressures relevant to commercial drying. This test series, referred to as a “dry” test series in this report, spans three uniform waterproof heater rod powers (representing spent fuel decay heats), four helium fill pressures, and six vacuum levels. This test series was conducted to cover the range of expected ADCS testing conditions for upcoming “wet” testing, where water will be introduced and a simulated commercial drying cycle will be performed. The dry test conditions were derived from the commercial drying conditions seen in the High Burnup Demonstration and the vacuum drying conditions chosen for a smaller scale Dashpot Drying Apparatus tested at Sandia National Laboratories in FY22. For a given uniform power and pressure/vacuum level, the ADCS was operated at constant power and pressure and allowed to reach steady state conditions. The thermal data obtained from these tests were analyzed, and the results can inform computational models built to simulate commercial drying processes by providing baseline thermal data prior to the introduction of water. Following the preliminary dry tests, a test plan for the ADCS will be developed to implement a drying procedure that begins with the introduction of water to the system and is based on measurements from the drying process used for the High Burnup Demonstration Project. While applying power to the simulated fuel rods, this procedure is expected to consist of filling the ADCS vessel with water, draining the water with applied pressure and multiple helium blowdowns, evacuating additional water with a vacuum drying sequence at successively lower pressures, and backfilling the vessel with helium. Additional investigations are expected to feature failed fuel rod simulators with engineered cladding defects and guide tubes with obstructed dashpots to challenge the drying system with multiple water retention sites. The data from these investigations is expected to inform the efficacy of commercial drying operations through the quantification of residual water in a prototypic-length dry storage canister.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

MARVEL Reactor Fuel Performance Report (Rev.2)

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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

MARVEL Reactor Fuel Performance Report

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.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Verification of the Coupled Neutronics/Thermal-Hydraulics Code TANSY against the MSFR Benchmark

The MSRs is a liquid-fuel reactor in which uranium, fission products and actinides are dissolved in the molten salt. The molten salt reactors are featured by circulating fuel, low pressure and high temperature. The MSRs meets many of the future goals of nuclear technology, in particular for what concerns as improved sustainability and unique characteristics in terms of actinide burning. The fluid nature of fuel in MSRs gives extra flexibility in reactor design, fuel fabrication and recycling. But because of the application of fluid fuel, MSRs have several special characteristics compared with the traditional solid-fuel reactors. Firstly, the transport of the delayed neutron precursors by the fluid flow reduces the contribution of delayed neutrons to the chain reaction.. Secondly, the fact that the fuel is dissolved in the coolant rather than separated from the coolant by the claddings results in a much stronger coupling phenomenon between the neutronics and thermal-hydraulics. Thirdly, the shape of the fuel is determined by the container because of the characteristics of fluid, the unstructured meshes are required to model the complex geometry accurately. However most of the conventional reactor codes were developed for solid-fuel reactor and therefore not capable of taking liquid fuel flow into account. To address all these particular modeling issues of MSRs, a multi-physics code named TANSY has been developed in our laboratory. In this paper, the benchmark problem based on the MSFR design was calculated to verify the code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Loss of Flow Conditions in a Modern Pool-type SFR and In-Pile Experiment Design

The metallic fuel safety performance under unprotected design-basis transients is a key consideration for the deployment of advanced sodium fast reactors (SFRs). Reliable data are needed to validate advanced safety codes, reduce uncertainty in cladding failure thresholds, and strengthen confidence in licensing approaches. To address this need, this report develops blueprints for a conceptual sodium loss-of-flow (LOF) experiment in the Mk-IIIR loop at the Transient Reactor Test Facility (TREAT), providing the technical foundation for future integral testing.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Progress on Design and Production of Oxide Dispersion–Strengthened Alumina-Forming Austenitic Alloys for Nuclear Applications

Alumina-forming austenitic (AFA) alloys are a promising class of nuclear materials due to their high-temperature oxidation/corrosion resistance and mechanical properties. Unfortunately, these alloys are limited in use for core material applications, specifically in their use as nuclear fuel cladding. Improving these alloys through a fine dispersion of oxide precipitates and thus increasing the effective irradiation sink strength of the alloy system may mitigate many of the degradation phenomena expected during alloy deployment (such as high-temperature helium embrittlement for lead-cooled fast reactor applications). This work package has started using a combination of conventional and advanced manufacturing approaches to the fabrication of oxide dispersion–strengthened (ODS) AFA materials. As such, this report summarizes progress to date in the fabrication of new compositions of these ODS AFA materials with a focus on reactor-specific design and a multifaceted manufacturing approach. Two different compositions of AFA gas-atomized powders were procured, and 200 g of each have been prepared using mechanical alloying with the expectation of high-temperature consolidation in Q1 of FY24. In addition, an approach to use reactive cover gases to promote additional precipitate formation has been developed for implementation in FY24 as a direct comparison with conventionally manufactured alloys.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS2005

Chemical Sensors Based on Optical Ring Resonators

Chemical sensors based on optical ring resonators are undergoing development. A ring resonator according to this concept is a closed-circuit dielectric optical waveguide. The outermost layer of this waveguide, analogous to the optical cladding layer on an optical fiber, is a made of a polymer that (1) has an index of refraction lower than that of the waveguide core and (2) absorbs chemicals from the surrounding air. The index of refraction of the polymer changes with the concentration of absorbed chemical( s). The resonator is designed to operate with relatively strong evanescent-wave coupling between the outer polymer layer and the electromagnetic field propagating along the waveguide core. By virtue of this coupling, the chemically induced change in index of refraction of the polymer causes a measurable shift in the resonance peaks of the ring. In a prototype that has been used to demonstrate the feasibility of this sensor concept, the ring resonator is a dielectric optical waveguide laid out along a closed path resembling a racetrack (see Figure 1). The prototype was fabricated on a silicon substrate by use of standard techniques of thermal oxidation, chemical vapor deposition, photolithography, etching, and spin coating. The prototype resonator waveguide features an inner cladding of SiO2, a core of SixNy, and a chemical-sensing outer cladding of ethyl cellulose. In addition to the ring Chemical sensors based on optical ring resonators are undergoing development. A ring resonator according to this concept is a closed-circuit dielectric optical waveguide. The outermost layer of this waveguide, analogous to the optical cladding layer on an optical fiber, is a made of a polymer that (1) has an index of refraction lower than that of the waveguide core and (2) absorbs chemicals from the surrounding air. The index of refraction of the polymer changes with the concentration of absorbed chemical( s). The resonator is designed to operate with relatively strong evanescent-wave coupling between the outer polymer layer and the electromagnetic field propagating along the waveguide core. By virtue of this coupling, the chemically induced change in index of refraction of the polymer causes a measurable shift in the resonance peaks of the ring. In a prototype that has been used to demonstrate the feasibility of this sensor concept, the ring resonator is a dielectric optical waveguide laid out along a closed path resembling a racetrack (see Figure 1). The prototype was fabricated on a silicon substrate by use of standard techniques of thermal oxidation, chemical vapor deposition, photolithography, etching, and spin coating. The prototype resonator waveguide features an inner cladding of SiO2, a core of SixNy, and a chemical-sensing outer cladding of ethyl cellulose. In addition to the ring res

Homer, Margie↗

NASA NTRS1972

Fabrication of fuel pin assemblies, phase 3

Five full size and eight reduced length fuel pins were fabricated for irradiation testing to evaluate design concepts for a fast spectrum lithium cooled compact space power reactor. These assemblies consisted of uranium mononitride fuel pellets encased in a T-111 (Ta-8W-2Hf) clad with a tungsten barrier separating fuel and clad. Fabrication procedures were fully qualified by process development and assembly qualification tests. Detailed specifications and procedures were written for the fabrication and assembly of prototype fuel pins.

Keeton, A. R.↗

NASA NTRS1979

Automated Array Assembly, Phase 2

The Automated Array Assembly Task, Phase 2 of the Low Cost Silicon Solar Array Project is a process development task. The contract provides for the fabrication of modules from large area tandem junction cells (TJC). During this quarter, effort was focused on the design of a large area, approximately 36 sq cm, TJC and process verification runs. The large area TJC design was optimized for minimum I squared R power losses. In the TJM activity, the cell-module interfaces were defined, module substrates were formed and heat treated and clad metal interconnect strips were fabricated.

Carbajal, B. G.↗

DOE OSTI2023

Irradiation performance of a U-7Mo in Al-Si matrix dispersion full-size fuel plate assembly

The Korea Atomic Energy Research Institute (KAERI) is leading the Ki-Jang Research Reactor (KJRR) project with the intent to develop a new reactor for medical isotope production and other nuclear research purposes. The KJRR core is designed to use high density fuel system where uranium alloyed with 7 wt% molybdenum (U-7Mo) particles are dispersed in a matrix of aluminum alloyed with 5 wt% silicon (Al-5Si) and clad in aluminum alloy 6061 (Al-6061) to form fuel plates. KAERI developed a fabrication facility to construct KJRR fuel assemblies and partnered with the Idaho National Laboratory (INL) to irradiate a full-size fuel assembly, with 21 total fuel plates, in the Advanced Test Reactor (ATR). Irradiation testing and subsequent Post Irradiation Exam (PIE) campaigns were performed successfully over a multi-year project. Monte Carlo neutronic calculations, coupling with a depletion code, were performed based on ATR’s as-run power history which showed that the highest power plate (plate 20) reached 83.1 % end-of-life (EOL) local burnup based on initial 235 U content. Additionally, finite element thermal modeling was performed based as-run power history which showed a beginning-of-life (BOL) peak local heat flux of 184 W/cm 2 . No anomalous fuel performance was observed during the irradiation and target test conditions were achieved. PIE showed favorable performance of the fuel assembly regarding all important phenomena. This paper describes the KJRR fuel assembly irradiation conditions and PIE data to support the conclusion that it performed well, without evidence of unexpected or problematic fuel performance, within an irradiation test designed to bound the KJRR design environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Wireless Instrumented RB Experiment Preliminary Design and Analysis

The ability to deploy new nuclear fuels for current or future reactor concepts requires carefully designed experiments to generate data to support fuel qualification. Ideally these experiments would include state of-the-art sensing to maximize the amount of in situ data that can be collected during operation. Furthermore, advanced reactor systems can take advantage of integrated in-core sensing technologies to maximize fuel utilization, reduce unnecessary conservativism in design margins, and improve operator’s understanding of limiting peaking factors. Before any novel sensing technologies can be readily adopted for nuclear applications, they must first demonstrate acceptable performance in test reactors. This report summarizes the preliminary design and analysis of the most highly instrumented irradiation experiment ever performed in the removable beryllium (RB) positions of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The Wireless Instrumented RB Experiment 2021 (WIRE-21) will test a wide range of sensors including wireless sensors being developed by Westinghouse Electric Company (WEC) that could provide in situ measurements of peak fuel temperatures and fuel rod pressurization due to fission gas release. The ability to wirelessly transmit a signal through the fuel rod’s cladding is critical to improving fuel monitoring capabilities without requiring signal penetrations through the cladding pressure boundary, which would significantly impact fuel fabrication, handling, and operation. Other sensors that will be tested in WIRE-21 include an array of thermocouples, self-powered neutron detectors (SPNDs), and spatially distributed fiber-optic temperature sensors. More generally, WIRE-21 will establish a flexible irradiation vehicle design to allow accelerated, economical testing of advanced sensor technologies while leveraging the extremely high neutron flux that is available in HFIR. This report summarizes the mechanical design for WIRE-21, the experimental test matrix, initial neutronic and thermal design analyses, and the active monitoring and control system enhancements necessary to support testing of advanced sensor technologies. The containment for WIRE-21 is similar to previous RB irradiation vehicles but includes a few modifications, most notably the use of integrated compression seals to pass a larger number of sensor leads through the experiment’s pressure boundary. In addition to the sensor leads, inert gas lines are passed into the experiment to enable active temperature control and the ability to pneumatically actuate a bellows-driven pressure sensor. WIRE-21 is targeting temperatures (300–350°C) and neutron fluence levels (~10 22 n/cm 2 ) relevant to light water reactors (LWRs), but the flexible design of the experiment vehicle allows much higher operating temperatures (>1,100°C). Neutronic calculations determine the neutron flux conditions as well as the nuclear heating within the experiments. These results are used as inputs to detailed thermal finite element calculations, which are required to evaluate the complex, three-dimensional heat transfer that occurs within WEC’s wireless sensor enclosures. Initial results show that the temperatures of the sensors’ enclosures and the metal bellows can be operated near the temperature range of LWR coolants and cladding while simultaneously increasing the temperature of a surrogate fuel material to values in the range of 800–1200°C to simulate centerline fuel temperatures during LWR operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

White Paper: Research & Development for the Time at Temperature Approach

Recent advancements in nuclear power research are greatly improving reactor safety and performance through the development of Accident Tolerant Fuel (ATF) and Low-Enriched Uranium Plus (LEU+). These innovations can address Departure from Nucleate Boiling (DNB) margins, which are vital for reactor safety. DNB happens when the coolant switches to film boiling, significantly decreasing heat transfer and posing a risk of fuel cladding failure. The U.S. Nuclear Regulatory Commission (NRC) employs conservative DNB criteria, which can potentially restrict the operational flexibility and efficiency of reactors. The Time at Temperature (TaT) approach could provide a more detailed and adaptable operational guideline by establishing acceptable time-temperature limits, accounting for the duration a material can withstand elevated temperatures without losing its integrity. This method allows reactors to operate more efficiently and safely, offering additional operational margins, faster power adjustments, and improved fuel cycle economics. TaT criteria allow for higher power levels and more flexible responses to operational transients, particularly applicable for anticipated operational occurrences (AOOs) that result in short durations of post-DNB conditions. It enhances plant operational flexibility, allows faster startup times, and enables quicker power level adjustments, optimizing fuel loading patterns and improving fuel cycle economics. Implementing TaT limits reduces core design constraints, lowers fuel usage, and reduces costs, essential for the long-term sustainability of Light Water Reactors (LWRs). TaT maximizes the use of advanced fuel technologies like ATF and LEU+, further enhancing their economic and environmental benefits. To apply the TaT approach in existing LWRs, collaborative research activities among various DOE-sponsored programs are essential. These efforts should incorporate fuel experiments, physics-based high-fidelity modeling, ML-based surrogate modeling, and optimization techniques. This whitepaper proposes four research and development areas: 1) Investigation of the feasibility of new operations of LWR with updated safety limits; 2) Assessment of reactor operation limits through uncertainty reduction; 3) Evaluation of power uprate in virtual environment; and 4) Lattice and reactor core design for power uprate. Each area includes why this research is in need and a suggested scope of work. These comprehensive research areas ensure practical and beneficial advancements for existing reactors, translating innovations in nuclear fuel and cladding technology into improved reactor performance and safety.

42 - ENGINEERING↗

DOE OSTI2018

Summary Technical Report for Conceptual Design of MARCH-SERTTA

The Transient Reactor Test facility (TREAT) was constructed in the late 1950’s, provided thousands of transient irradiations before being placed in standby in 1994, and resumed reactor operations in 2017 in order to reclaim its crucial role in nuclear-heated safety research [1]. Starting in September of 2018, TREAT performed its first series of new fueled transient experiments using a novel modular irradiation vehicle, termed the Minimal Activation Retrievable Capsule Holder (MARCH) system. The MARCH system was developed to support high throughput testing and post transient examination by using fresh fuel samples arranged in low-activation hardware materials. Deployment of this irradiation vehicle system was supported by the Accident Tolerant Fuels (ATF) program. These initial experiments used inert gas atmosphere capsules (referred to as the Separate Effects Test Holder, SETH) containing small rodlets composed of UO2 pellets in zirconium alloy cladding with temperature measurement instruments. The initial SETH campaign was conducted in order to commission irradiation capabilities and deepen the understanding of specimen to core energy coupling. Building upon the successes of the initial SETH campaign, an evolution of the capsule design has been conceptualized in order to support the same kind of high throughput, cost effective, rodlet scale testing, but with the notable improvement of being able to immerse specimens in liquid water for testing conditions better representing Light Water Reactors (LWR). This capsule is referred to as the MARCH system Static Environment Rodlet Transient Test Apparatus (MARCH-SERTTA).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Continuous wave room temperature operation of the 2 $μ$m GaSb-based photonic crystal surface emitting diode lasers

Continuous wave room temperature operation of 2 μm GaSb-based photonic-crystal surface-emitting diode lasers has been realized. The deep etched square mesa devices showed threshold current densities of 500 A/cm 2 at 20 °C. The epi-side down mounted lasers generated above 10 mW of output power in the continuous wave regime and tens of milliwatts in pulses from the 200 × 200 μm 2 aperture. Here, the breakthrough in the device performance parameters was achieved thanks to a highly homogeneous air-pocket retaining epitaxial regrowth process optimized for a specifically designed antimonide diode laser heterostructure. The nanofabrication method utilizing low temperature atomic hydrogen surface cleaning yielded low disorder square lattice of droplet-shaped voids covered by uniform p-cladding layer. The laser emission spectrum as well as near/far field patterns demonstrated peculiar features presumably linked to deformation of the void shape during regrowth and formation of the array of filaments.

77 NANOSCIENCE AND NANOTECHNOLOGY↗

DOE OSTI2017

FY2017 Updates to the SAS4A/SASSYS-1 Safety Analysis Code

The SAS4A/SASSYS-1 safety analysis software is used to perform deterministic analysis of anticipated events as well as design-basis and beyond-design-basis accidents for advanced fast reactors. It plays a central role in the analysis of U.S. DOE conceptual designs, proposed test and demonstration reactors, and in domestic and international collaborations. This report summarizes the code development activities that have taken place during FY2017. Extensions to the void and cladding reactivity feedback models have been implemented, and Control System capabilities have been improved through a new virtual data acquisition system for plant state variables and an additional Block Signal for a variable lag compensator to represent reactivity feedback for novel shutdown devices. Current code development and maintenance needs are also summarized in three key areas: software quality assurance, modeling improvements, and maintenance of related tools. With ongoing support, SAS4A/SASSYS-1 can continue to fulfill its growing role in fast reactor safety analysis and help solidify DOE’s leadership role in fast reactor safety both domestically and in international collaborations.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2023

Preliminary Simulations of Commercial Drying Cycles Using the Advanced Drying Cycle Simulator

The purpose of this report is to document updates on the apparatus to simulate commercial vacuum drying procedures at the Nuclear Energy Work Complex at Sandia National Laboratories. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system during subsequent storage and disposal. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates well-designed investigations of drying process efficacy and water retention. Scaled tests that incorporate relevant physics and well-controlled boundary conditions are essential to provide insight and guidance to the simulation of prototypic systems undergoing drying processes. This report documents fiscal year 2023 (FY23) updates on the Advanced Drying Cycle Simulator (ADCS). This apparatus was built to simulate commercial drying procedures and quantify the amount of residual water remaining in a pressurized water reactor (PWR) fuel assembly after drying. The ADCS was constructed with a prototypic 17×17 PWR fuel skeleton and waterproof heater rods to simulate decay heat. These waterproof heaters are the next generation design to heater rods developed and tested at Sandia National Laboratories in FY20. In FY23, a series of four simulated commercial drying tests was completed. This report presents the temperature and pressure histories of the drying tests as well as axial temperature profiles that can be compared to data from the Electric Power Research Institute (EPRI) High Burnup Demonstration TN-32B cask. Water content measurements and dew point calculations from a Hiden Analytical HPR-30 mass spectrometer are also presented in this report. Due to familiarization with this first-of-a-kind system, refinements to equipment calibration and test procedures have been identified to better match commercial drying cycles for future simulated tests. However, the presented data demonstrate the successful construction and operation of a viable research platform for quantifying residual water content closely approaching that expected in dry storage canisters during commercial drying procedures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗