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NASA NTRS2019

Hybrid Fuel Coupling in a Pulsed Z-Pinch Rocket Engine

The work presented here sought to explore a portion of the parameter space of a hybrid nuclear fuel in regards to ignition and burn by analyzing the effect of initial geometry and thermodynamic conditions. The authors performed 0D power balance and 1D burn wave calculations to determine temperature progression and energy production for defined initial conditions. Geometries examined are representative of concept fuels for a Pulsed Fission-Fusion (PuFF) engine. This work focuses on lithium deuteride and uranium 235 for the fuel since these are seen as leading candidates for PuFF. Presented below is a power balance illustrating a reduction in the energy and density required to breakeven of hybrid fuels in comparison with fusion fuels. Also the impact of fusion and fissile fuel quantities upon initial energies is presented. One can see that the initial energy required to breakeven in a hybrid cylindrical nuclear fuel decreases with decreasing fissile liner thickness, decreasing fusion fuel core radius, and increasing compression ratio of the fusion fuel.

Taylor, Brian↗

NASA NTRS2019

Hybrid Fuel Coupling in a Pulsed Z-Pinch Rocket Engine

Taylor, Brian↗

DOE OSTI2023

Numerical Analysis of Novel Plate Type Heat Exchanger with Oval-Twisted Channels

The novel heat exchanger (HX) designs implementing geometrical and surface modification combinations are expected to perform better than traditional HX technologies. Applied enhancement techniques seek to achieve (1) higher overall heat transfer performance, (2) increased compactness, and (3) simplified or comparable manufacturability. One innovative enhancement technique is to generate swirling flow vortices induced by channel or tube twisting. The turbulence generated by the twisted cross-section greatly enhances the heat transfer rate with minimal increases in pressure drop. Plate-type HXs and Printed Circuit Heat Exchangers (PCHXs) are compact designs that achieve high heat transfer rates per unit volume by utilizing several small channels, which maximizes the heat transfer surface area between the hot and cold fluids. Currently, advanced manufacturing technologies enable the design and fabrication of compact-type units with complicated channel geometries to achieve the highest performance and meet the compactness criteria of innovative HX technology. The proposed HX design concept combines the compactness of plate-type HXs and twisted channels, which provide additional turbulence and flow swirl enhancement. The plate-type oval-twisted HX (PTOTHX) is a crossflow configuration, with 16 short channels on one side (for hot fluid) and 8 long channels on the perpendicular side (for cold fluid). The inlet plenums have flow guide vanes to redirect flow and produce uniformity across the various flow paths. The compact size and purportedly improved heat transfer performance of the PTOTHX investigated herein prove its viability in various applications. Some notable potential nuclear applications of the PTOTHX include reactor core, spent fuel cooling, and residual heat dissipation. To establish a reference case, circular channels (PTCHX) are also considered in the present study for comparison with (PTOTHX). This paper aims to outline the numerical analysis procedures for determining the viability of the PTOTHX by comparing its heat transfer performance with the PTCHX units. The computational study used STAR-CCM+, a commercial computational fluid dynamics (CFD) code. Sensitivity analysis and model selection studies are conducted to determine the appropriate mesh density and turbulence model to provide the reported results. Numerical analyses comparing the Nusselt number (Nu) of the PTOTHX design with a comparable HX unit, including a circular cross-section and no twisting (PTCHX), show an overall heat transfer performance increase of 29-55% for balanced flow and 29-59% for imbalanced flow. Oval-cross-sectional twisted channels induce swirling flow vortices, enhancing the working fluid's convective heat transfer capabilities.

25 ENERGY STORAGE↗

DOE OSTI2024

Wrought FeCrAl alloy (C26M) cladding behavior and burst under simulated loss-of-coolant accident conditions

Cladding burst experiments for FeCrAl cladding were performed in the Severe Accident Test Station facility at Oak Ridge National Laboratory. These experiments were simulated using the BISON fuel performance code to better understand the cladding plastic behavior and failure under simulated loss-of-coolant accident conditions. 3D cladding surface boundary conditions were generated using composite axial and azimuthal profiles from experiment thermocouple data. To improve the simulation analysis capabilities in BISON for cladding burst behavior, new thermal creep, plasticity, and failure stress models specific to C26M, a wrought FeCrAl alloy, were developed and implemented. Initial cladding burst results indicated a general underprediction in the failure temperature of the six cladding burst simulations versus the observed failure temperatures. Close investigation of the experiment timing versus the underlying tensile test data revealed that, compared with the tensile specimens, the cladding tubes did not experience the same long holding time at high temperatures. New tensile tests were performed at high temperatures using a temperature ramp similar to the simulated loss-of-coolant accident experiments. These new tensile curves showed an approximately 80% increase in the ultimate tensile strength of the C26M alloy, indicating that a holding time of 10 min at 700 °C and 800 °C allows annealing to change the material microstructure. Using the updated tensile properties, the burst temperatures and stresses from the simulations showed remarkable agreement with the experimental results. This study was then extended by varying the initial pressure to highlight the burst temperature difference between standard Zircaloy-4 and C26M cladding under equivalent conditions. The results show that C26M has a burst temperature that is approximately 70–130 K greater than that of Zircaloy-4. In conclusion, these modeling predictions can be further improved by collecting high-temperature tensile data for C26M beyond the temperature ranges used in this work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Assembly of MiniFuel Targets for Irradiation of TRISO Fuel Compacts in the High Flux Isotope Reactor

To support the development of Kairos Power’s fluoride-salt-cooled high-temperature reactor, irradiation testing of tristructural isotropic (TRISO) fuel compacts was performed at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) to collect experimental data on TRISO fuel during high particle-power operation and validate fuel performance models. Fuel compacts containing enriched uranium oxycarbide (UCO), natural UCO (NUCO), or uranium dioxide (UO 2 ) TRISO particles were fabricated at ORNL and inserted into MiniFuel targets for HFIR irradiation. Five MiniFuel targets were successfully assembled, welded, tested, and delivered to HFIR, along with their quality assurance documentation. The targets were inserted into HFIR’s inner vertical experiment facility within the permanent beryllium reflector. Each target contains six fuel compacts and will be irradiated in HFIR for four cycles, with target temperatures of 500, 700, and 900°C. This report summarizes the experiment design, test matrix, and fabrication. This work was performed under the Nuclear Science User Facility program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Non-Equilibrium Actinide Radiation Chemistry and the Nuclear Fuel Cycle

Actinides are inherently unstable elements that frequently coexist with other radioisotopes, generating intense ionizing radiation fields that drive the formation of non equilibrium oxidation states. These transient species exert a profound mechanistic influence on the radiation response of actinide containing systems due to their unique redox chemistry. Despite their importance, they remain poorly understood, yet such insight is essential for advancing actinide science and accurately predicting radiation driven behavior. Actinide separations—critical for nuclear energy technologies, strategic deterrence, space exploration, and nuclear medicine—depend on precise control of actinide oxidation states to recover targeted elements from complex matrices such as used nuclear fuel. However, during these processes, actinides, their coordination complexes, and the separation media are all exposed to intense, multicomponent (alpha, beta, gamma, etc.) radiation fields that can alter process efficiency, selectivity, and chemical stability. Understanding, controlling, and mitigating radiation induced reactions is therefore key to innovating and optimizing next generation separation technologies. This seminar will provide an overview of the nuclear fuel cycle and non equilibrium actinide radiation chemistry in the context of recovering actinides from used nuclear fuel, with a particular emphasis on direct dissolution–based reprocessing strategies. We will explore time resolved electron pulse radiolysis and alpha and gamma dose accumulation studies, integrated with multiscale computational modeling, to elucidate the molecular level roles of radiation driven, non equilibrium actinide species in process performance and in the radiolytic stability of organic ligands used for actinide recovery. These insights offer new pathways for designing advanced separation methods and next generation solvent systems, with broad implications for the future of the nuclear fuel cycle.

37 - INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL C↗

DOE OSTI2023

ECAR-7210 Rev 0 Steady-state Thermo-mechanical Analysis of the MARVEL Fuel Bundle

This document reports the steady-state thermal-hydraulic and static mechanical analyses of the MARVEL micro-reactor core. The objective of this work is to investigate whether fuel-to-fuel contact could occur due to the thermal deformation of the fuel cladding during the normal operation and to assess the resulting impact on the peak cladding temperatures (PCT) if contact occurs. The scope of the work consists of three computational tasks. First, a conjugate heat transfer analysis using computational fluid dynamics (CFD) model was performed to evaluate the PCTs during normal operating conditions. Second, a finite element analysis (FEA) using the cladding temperature field obtained from the CFD analysis was performed to evaluate the thermal deformation of cladding and determine the occurrence of fuel-to-fuel contact. Finally, a CFD analysis of FEA-informed fuel-to-fuel gap (represented by a conservative 0.05 mm uniform gap) was performed to evaluate whether the PCT exceeds the safety criteria.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

DOE OSTI2021

Metallic fuel cladding degradation model development and evaluation for BISON

Fuel cladding chemical interaction (FCCI) and coolant cladding chemical interaction (CCCI), both consume stainless-steel cladding in sodium-cooled fast reactors (SFRs), and effectively degrade cladding mechanical performance through thickness reduction. Therefore, FCCI/CCCI and corresponding cladding degradation models are essential for advanced fuel performance codes, such as BISON, to reliably predict cladding behavior in SFRs. Here, we report the development and implementation efforts for BISON's FCCI/CCCI correlations and cladding degradation models based on U.S. legacy metallic fuel data. The models were evaluated using Integral Fast Reactor (IFR) program X447/A experiment data supported by the ongoing integration project enabling coordinated application of BISON and EBR-II fuel irradiation and physics database (FIPD). Furthermore, the implemented models were demonstrated to improve BISON's capabilities of predicting cladding damage and degradation behavior that is more consistent with post-irradiation examination observations. Additionally, some limitations of current BISON modules are identified, which are to be overcome through the BISON-FIPD integration efforts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Sub-critical assembly die-away analysis with organic scintillators

Determining the reactivity of systems containing fissile material is essential for verifying safety, and possession and enrichment levels. The goal of this experiment was to characterize a subcritical nuclear fuel assembly irradiated by a pulsed neutron generator through differential die-away analysis. Similar characterizations are commonly performed with thermal neutron detectors, but these detectors are bulky and cumbersome, the materials are not widely available, and timing can be clouded by the required thermalization time. In this work, an organic scintillator capable of detecting both neutrons and photons was used to measure the die-away profiles of both particles. These time profiles were then used to estimate the reactivity of the assembly. Here, the calculated reactivity estimates compare favorably to simulated estimates, and therefore provide a useful tool for fissile material characterization.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2025

Multilayer Architecture Barriers for High Temperature Nuclear Applications

Innovative thin film multilayer metal/ceramic barrier coatings have been developed by Argonne National Laboratory to protect fuel cladding against different phenomena that can be encountered under the harsh operating conditions of advanced nuclear reactors. One phenomenon is Fuel-Cladding Chemical Interaction (FCCI), which can represent a challenge to the performance of steel-clad metallic fuels in sodium-cooled fast reactors (SFRs), especially when operating to high burnups at elevated cladding temperatures, and long fuel residence time. Another phenomenon is the corrosion of zirconium-based cladding under prototypic boiling water reactor (BWR) conditions and associated cladding during off normal conditions, including loss of coolant accidents (LOCA). This work provides details of the development and testing of a thin film (few microns) multilayer metal/ceramic (CrY/Y 2 O 3 ) coatings on steel cladding and its performance against FCCI. The coating architecture evolved through systematic mechanical testing, radiation tolerance assessments, and high-temperature diffusion studies from a preliminary multilayer concept to an optimized high-performance design. The optimized structure leverages stoichiometric Y 2 O 3 for its exceptional radiation stability, chemical inertness, and ultra-low fission product diffusivity (e.g., Ce diffusivity ~10 -20 m 2 /s in irradiated Y 2 O 3 at 650 °C). This is paired with ductile chromium-rich interlayers (Cr 95 Y 5 ), which significantly enhance fracture toughness and mechanical resilience. Design refinements, including increased ceramic layer thickness and optimized Cr–Y interlayer composition, eliminated fracture propagation and preserved coating integrity under extreme mechanical stress (>10× typical reactor conditions) and prolonged radiation exposure (~300 dpa). Those conditions are well beyond expected conditions during normal operations in SFR, and cover some of the transient conditions as well. In addition, the coating deposition was demonstrated to be applied uniformly to the inner surface of steel tubes representing prototypic SFR cladding. Another phenomenon under consideration in this work is the corrosion of zirconium-based cladding under prototypic boiling water reactor (BWR) conditions and associated off normal conditions including loss of coolant accident (LOCA). Another multilayer thin film coating is considered for mitigation of this phenomenon, where a coating of yttrium aluminum and yttrium aluminum ceramic/metal thin film coating. Preliminary results on the coating performance during high temperature thermal cycling (up to 1000 C) have shown promising performance of the coating. On-going optimization of the coating will continue in FY26.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

SAS4A/SASSYS-1 Version 5.7 and 5.6 Release Notes

SAS4A/SASSYS-1 (SAS) is a simulation tool used to perform deterministic analyses of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. In this report, the major developments that were implemented in Versions 5.7 and 5.6 are summarized and an overview of all changes made between versions is presented. SAS Version 5.7 and 5.6 introduced a number of modeling improvements that allow users to capture important phenomena, such as pump heating and metallic fuel performance.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Effects of Internal and External Heat Sources on Cladding Microstructure and Rupture Performance

In the event of a Loss of Coolant Accident (LOCA), the primary supply of cooling water for a nuclear reactor is lost, leading to a significant pressure differential across the cladding wall. Without adequate cooling, the fuel rods continue to heat as a result of fission reactions. Research at Oak Ridge National Laboratory’s (ORNL’s) Severe Accident Test Station (SATS) is currently focused on evaluating fuel cladding performance using an external infrared lamp as a heat source, whereas legacy testing primarily utilized an internal heating approach. While external heating may better simulate the effect of neighboring fuel rods heating a central rod, it may not accurately represent the internal heat absorption from the fuel during accident transients. The heating dynamics depend greatly on the fuel rod's position within the bundle and the reactor. To fully understand the implications of a LOCA event and assess the influence of internal heating on cladding performance, combined internal heating and pressurization capability was developed at ORNL. Tests were conducted to compare cladding segments heated internally (representing heat from the fuel within the rod) and externally (representing heat from adjacent fuel rods). The findings indicate that both internal and external heating result in comparable rupture temperatures during 5°C/s laboratory LOCA tests and also agree with legacy test data. Axial temperature gradients and internal heat source dispersal were found to significantly impact cladding deformation and rupture geometry. Clear modifications were outlined to further improve the capability with heating rates above 5°C/s.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Effects of Internal and External Heat Sources on Cladding Microstructure and Rupture Performance

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Analysis with the ORIGEN Module of PNAR Spent Fuel Measurements for Nuclear Safeguards Applications

This report documents the analysis of the Passive Neutron Albedo Reactivity (PNAR) measurements for 23 boiling water reactor spent fuel assemblies that were performed in Finland under Action Sheet 65, which is an international collaboration on spent fuel safeguards verification methods in the context of the Finnish spent fuel encapsulation/repository system. PNAR measures the passive neutron and gamma emission rates from each of the spent fuel assemblies like a Fork detector, and it also measures the PNAR ratio, which is expected to correlate with the net neutron multiplication of the measured fuel assembly. The analysis was performed with the Oak Ridge Isotope Generation and Depletion (ORIGEN) Data Analysis Module, which was originally developed for predicting Fork detector spent fuel measurement signals in real time and has been integrated into the Integrated Review and Analysis Package developed by Euratom and the International Atomic Energy Agency. The module includes the ORIGEN burnup analysis code and integrates detector response functions pregenerated using the MCNP Monte Carlo transport code to predict the detector signals in several seconds per assembly. In this study, new response functions specific to the analyzed PNAR measurements were generated for the ORIGEN Module. The study also analyzes the impact of using detailed fuel design and operation information vs. standard safeguards information on the results calculated with the ORIGEN Module. Using detailed information reduced the standard deviation of the relative differences between calculated and measured neutron count rates among the 23 assemblies from ~10% to ~4%. The results obtained using standard safeguards information for these PNAR measurements were similar to those obtained for the Fork detector. A clear trend was found between the calculated net neutron multiplications and the measured PNAR ratios of the 23 assemblies. An uncertainty assessment of the calculations was also performed to estimate the potential impact of the accuracy and the completeness of declaration information on the calculated results for the count rates and net neutron multiplications. The ORIGEN Module predictions for PNAR signals and net neutron multiplication are expected to directly support the safeguards inspectors’ efforts to verify operator declarations of a spent fuel assembly in real time in safeguards practices.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1963

Performance Capability of Single-Cavity Vortex Gaseous Nuclear Rockets

An analysis was made to determine the maximum powerplant thrust-to-weight ratio possible with a single-cavity vortex gaseous reactor in which all the hydrogen propellant must diffuse through a fuel-rich region. An assumed radial temperature profile was used to represent conduction, convection, and radiation heat-transfer effects. The effect of hydrogen property changes due to dissociation and ionization was taken into account in a hydrodynamic computer program. It is shown that, even for extremely optimistic assumptions of reactor criticality and operating conditions, such a system is limited to reactor thrust-to-weight ratios of about 1.2 x 10(exp -3) for laminar flow. For turbulent flow, the maximum thrust-to-weight ratio is less than 10(exp -3). These low thrusts result from the fact that the hydrogen flow rate is limited by the diffusion process. The performance of a gas-core system with a specific impulse of 3000 seconds and a powerplant thrust-to-weight ratio of 10(exp -2) is shown to be equivalent to that of a 1000-second advanced solid-core system. It is therefore concluded that a single-cavity vortex gaseous reactor in which all the hydrogen must diffuse through the nuclear fuel is a low-thrust device and offers no improvement over a solid-core nuclear-rocket engine. To achieve higher thrust, additional hydrogen flow must be introduced in such a manner that it will by-pass the nuclear fuel. Obviously, such flow must be heated by thermal radiation. An illustrative model of a single-cavity vortex system employing supplementary flow of hydrogen through the core region is briefly examined. Such a system appears capable of thrust-to-weight ratios of approximately 1 to 10. For a high-impulse engine, this capability would be a considerable improvement over solid-core performance. Limits imposed by thermal radiation heat transfer to cavity walls are acknowledged but not evaluated. Alternate vortex concepts that employ many parallel vortices to achieve higher hydrogen flow rates offer the possibility of sufficiently high thrust-to-weight ratios, if they are not limited by short thermal-radiation path lengths.

Ragsdale, Robert G.↗

DOE OSTI2026

Fuel Performance Modeling Internship Final Presentation

This study examines the performance of U-Zr and U-Pd-Zr annular metallic fuel rodlets and details the current status of modeling efforts regarding U-Pu-Zr solid metallic fuel rodlets which were experimentally evaluated as part of the Advanced Fuels Campaign (AFC). The AFC mission is to develop novel fuel technologies and facilitate the implementation of those technologies by industry partners. A key objective is to improve steady-state and transient performance over current fuel types. The experiments of interest in this study included metallic fuel rodlets within HT-9 cladding which were placed in SS-316 capsules and inserted in the Advanced Test Reactor (ATR). Certain mechanical and thermal conditions cannot be directly evaluated through experiments and fuel performance modeling is used to shed light on this evolution over time. In this study, BISON Multiphysics simulations are leveraged to investigate the state of the fuel system throughout and after the experimental conditions.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS2019

Near-Term Options for a Nuclear Thermal Propulsion Flight Demonstrator

The Appropriations Bill passed by the US Congress in February 2019 instructed NASA to direct “not less than $100,000,000 for the development of nuclear thermal propulsion, of which not less than $70,000,000 shall be for the design of a flight demonstration by 2024 for which a multi-year plan is required by both the House and the Senate within 180 days of enactment of this agreement." As part of NASA’s response to this direction, the Advanced Concepts Office (ACO) at the Marshall Space Flight Center (MSFC) was tasked with leading a study to develop a nuclear thermal propulsion (NTP) flight demonstration (FD) concept and evaluate its feasibility with respect to the near-term schedule goal. During formulation for the NTP FD study, two perspectives emerged with regards to FD concept design. The first seeks to strictly observe the immediate near-term schedule goal, embracing a completely “off-the-shelf,” high-TRL approach to subsystem design and component selection. The downside to this approach is that the propulsion performance to be expected from such a design is significantly lower than what NTP promises for operational systems, and the value of the flight demo is potentially reduced due to a lack of traceability. The second approach advocates for an FD concept that shows increased traceability to the projected designs of operational systems, providing risk reduction for future NTP-enabled missions. This option comes at the cost of schedule and development risks, as it requires some new investments in nuclear reactor fuels and design. In order to understand the implications and differences between these two approaches, the ACO team elected to perform a concept design of each type, labeling the immediate near-term concept Flight Demo 1 (FD1), and the higher traceability concept Flight Demo 2 (FD2). This paper will present a summary of the mission profiles and system designs for both FD1 and FD2, identifying key drivers and challenges for each design.

Edwards, S. J.↗

DOE OSTI2024

Characterization of spent nuclear fuel canister surface roughness using surface replicating molds

In this study we present a replication method to determine surface roughness and to identify surface features when a sample cannot be directly analyzed by conventional techniques. As a demonstration, this method was applied to an unused spent nuclear fuel dry storage canister to determine variation across different surface features. In this study, an initial material down-selection was performed to determine the best molding agent and determined that non-modified Polytek PlatSil23-75 provided the most accurate representation of the surface while providing good usability. Other materials that were considered include Polygel Brush-On 35 polyurethane rubber (with and without Pol-ease 2300 release agent), Polytek PlatSil73-25 silicone rubber (with and without PlatThix thickening agent and Pol-ease 2300 release agent), and Express STD vinylpolysiloxane impression putty. The ability of PlatSil73-25 to create an accurate surface replica was evaluated by creating surface molds of several locations on surface roughness standards representing ISO grade surfaces N 3 , N 5 , N 7 , and N 8 . Overall, the molds were able to accurately reproduce the expected roughness average (R a ) values, but systematically over-estimated the peak-valley maximum roughness (R z ) values. Using a 3D printed sample cell, several locations across the stainless steel spent nuclear fuel canister were sampled to determine the surface roughness. These measurements provided information regarding variability in normal surface roughness across the canister as well as a detailed evaluation on specific surface features (e.g., welds, grind marks, etc.). The results of these measurements can support development of dry storage canister ageing management programs, as surface roughness is an important factor for surface dust deposition and accumulation. This method can be applied more broadly to different surfaces beyond stainless steel to provide rapid, accurate surface replications for analytical evaluation by profilometry.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗