Idaho National Laboratory Research Articles

Post-irradiation examination of UN-Mo-W fuels for space nuclear propulsion

The National Aeronautics and Space Administration's return to space nuclear propulsion stems from the need for a more efficient method of space travel. Nuclear thermal propulsion systems have been shown to be two times more efficient than chemical propulsion. NASA's Sirius program was created to fabricate and test fuels for space nuclear propulsion, specifically to determine their performance under prototypical startup conditions. The Sirius project featured 4 test capsules, Sirius-1 featured uranium nitride fuel dispersed in a matrix of tungsten and rhenium, while Sirius-2A, -2B, and -3 featured uranium nitride-molybdenum-tungsten fuel (UN-Mo-W). This study discusses the Sirius-2A and -2B irradiation experiments at the Idaho National Laboratory, specifically their performance under irradiation at the Transient Reactor Test Facility. It was found that the fuel samples overall did not exhibit significant cracking, though the Sirius-2A fuel did have one large crack on the surface of the fuel. There was minimal hydrogen absorption in the samples, though it is unknown if the absorption occurred during irradiation or during fabrication. Mechanical testing indicated that the UN fuel demonstrated ceramic behavior as expected, and the Mo/W matrix demonstrated linear elastic behavior to failure.

Early-stage Testing of Thermal Power Dispatch Simulator for Subjective Mental Workload and Evolution Time

The excess thermal energy produced by nuclear power plants (NPPs) during low electricity demands can be utilized in industrial processes, such as hydrogen production, through a thermal power dispatch (TPD) system. Initial testing of the first iteration of a single-train TPD design with a manual control mode at the Idaho National Lab (INL) revealed a high operator workload and degraded control capability. The current study evaluated the impact of an enhanced dual-train TPD design on operators’ subjective mental workload and evolution task-time while completing two operating scenarios in manual and automatic control modes. The results showed no statistically significant difference between participants’ mental workload using both control modes. Evolution time in automatic control mode took a shorter time than in manual control, with participants completing all evolutions in less than the 10-min set as the design specification limit. The shorter evolution time is discussed within the context of plant safety and operational efficiency.

Using the National Land Cover Database as an indicator of shrub-steppe habitat: comparing two large United States federal lands with surrounding regions

ABSTRACT There is a need to assess whether ecological resources are being protected on large, federal lands. The aim of this study was to present a methodology which consistently and transparently determines whether two large Department of Energy (U.S. DOE) facilities have protected valuable ecological lands on their sites compared to the surrounding region. The National Land Cover Database (2019) was used to examine the % shrub-scrub (shrub-steppe) and other habitats on the DOE’s Hanford Site (HS, Washington) and on the Idaho National Laboratory (INL), compared to a 10-km and 30-km diameter band of land surrounding each site. On both sites, over 95% is in shrub-scrub or grassland, compared to the surrounding region. Approximately 70% of 10 km and 30-km bands around INL, and less than 50% of land surrounding HS is located in these two habitat types. INL has preserved a significantly higher % shrub/scrub habitat than HS, but INL allows grazing on 60% of its land. HS has preserved a significantly higher % grassland than INL but no grazing on site is present. The methodology presented may be used to compare key ecological habitat types such as grasslands, forest, and desert among sites in different parts of the country. This methodology enables managers, resource trustees, and the public to (1) make remediation decisions that protect resources, (2) assess whether landowners and managers have adequately characterized and protected environmental resources on their sites, and (3) whether landowners and managers have protected the integrity of that land as well as its climax vegetation.

Design of experimental facility for helium gas component testing

The current state-of-the-art helium gas test loops for component testing are constrained by the maximum operating temperatures, pressures, and flow rates of such facilities. Idaho National Laboratory is designing a new experimental facility known as HECTOR (HElium Component Testing Out-of-pile Research), which is a helium gas test loop for evaluating the performance of components to helium gas pressures, temperatures, and flow rates up to 8 MPa, 800 °C, and 0.15 kg/s, respectively. The construction of this facility will enable researchers to perform experiments across a wide range of Reynolds and Nusselt numbers, which could result in the maturation of technology for high-temperature gas-cooled reactor (HTGR) components.

The MOOSE fluid properties module

The Fluid Properties module within the Multiphysics Object-Oriented Simulation Environment (MOOSE) is used to compute fluid properties for numerous applications, ranging from nuclear reactor thermal hydraulics to geothermal energy. Those applications drove the development of the module to enable numerous different fluid equations of states, property lookups with primitive and conserved flow variable to cater to pressure and density-driven solvers, and an object-oriented design facilitating expansion and maintenance. Each fluid property is implemented in its own class but inherits capabilities such as automatic differentiation, automated out-of-bounds handling or variable conversion capabilities. This paper presents the module, its design, its user and developer interface, its content in terms of fluids and properties, and several of its applications showing its major role in the MOOSE simulation ecosystem. Program summaryProgram title: MOOSE Fluid Properties moduleCPC Library link to program files:https://doi.org/10.17632/cwzhsyp6pd.1Developer's repository link:https://github.com/idaholab/moose/tree/next/modules/fluid_propertiesLicensing provisions: LGPLProgramming language: C++, PythonNature of problem: The simulation of thermal hydraulics of advanced nuclear reactor systems, such as heat pipe micro-reactors or molten-salt cooled pebble bed reactors, requires a wide variety of discretizations of the fluid flow equations, from 1D thermal hydraulics to computational fluid dynamics at various levels of fidelity, with a wide variety of coolants. Applications are developed within the MOOSE C++ framework by Argonne and Idaho National Laboratories to simulate these reactors for research and design purposes. These applications (Sockeye, SAM, others) rely on MOOSE for the computation of fluid properties. The fluid properties module contains properties for most advanced nuclear reactor coolants, including an interface to the Molten Salt Thermodynamics Database (MSTDB) developed by Oak Ridge National Laboratory. Single phase, two phase, and gas mixtures fluid properties are computed by the module.Solution method: The fluid properties module includes numerous numerical methods to support the wide range of applications, notably forward automatic differentiation, conversion methods between pressure and density-driven variable sets, spline-based table lookups which are the current state of the art for the fast computation of fluid properties. The integration with MOOSE facilitates uncertainty quantification with regards to the fluid properties and optimization studies with regards to the fluid composition.

A Demonstration of the Risk-Informed NEI 18-04 Design Evaluation Model for the Modular High-Temperature Gas Reactor

Several advanced reactor designs are now under active development in the United States and elsewhere, promising sustainable solutions to the growing world energy needs. The designs currently being considered are quite diverse and different from the more established light water reactor technology that has dominated the operating commercial nuclear landscape. While advanced reactor concepts were first explored in the dawn of the nuclear age, they are now being reconsidered under the light of modern needs, and specifically, for their flexible operating conditions and inherent safety characteristics. In response, the U.S. Nuclear Regulatory Commission staff is moving forward with development of 10 CFR Part 53 rulemaking, which is a more risk-informed, technology-agnostic framework for licensing and regulating such new designs. The nuclear industry response to this regulatory initiative resulted in the technical report by the Nuclear Energy Institute, NEI 18-04 Revision 1, which provides an implementation roadmap of the risk-informed approach when defining the safety case for a new plant design. The implementation of this safety case may be a nontrivial exercise for an actual reactor design. This paper provides a demonstration of performing such an analysis for a representative advanced reactor. Public information from the General Atomics high-temperature gas reactor design was considered in this demonstration. The analysis workflow was facilitated with the FPoliSolutions’ proprietary Risk-Informed System Engineering (RISE) digital platform, a product that was presented in previous publications. RISE is one application of FPoli’s enterprise digital platform, which was created to facilitate orchestration of complex workflows leveraging recent technologies developed at national laboratories, such as Idaho National Laboratory’s RAVEN and EMRALD frameworks. The analysis described in the paper includes the selection and classifications of events, the integration of probabilistic risk analysis artifacts, and event modeling simulations for consequence evaluations. The results are then used for system, structures, and components safety classification and a synthesis of the safety case for the design in line with the frequency-consequence targets presented in NEI 18-04. The purpose of the analysis, as framed in RISE, is to readily produce outputs and views that can aid users and regulators in making risk-informed decisions to demonstrate their plant safety case.

Progress on Pu-238 production at Idaho National Laboratory from February 2022 to July 2023

Idaho National Laboratory (INL) has continued to qualify irradiation positions in the Advanced Test Reactor (ATR) for Pu-238 production to support NASA deep space missions. Over the past year, INL qualified Np-237 targets for ATR’s North East Flux Trap (NEFT), Inner-A and H positions. Work has begun to requalify the South Flux Trap (SFT) and to qualify the East Flux Trap (EFT) for the ATR GEN I target and is midway through the qualification process. This paper gives an overview of operational and technical activities from February 2022 to July 2023.

Post irradiation examinations of FeCrAl cladding in PWR conditions

FeCrAl alloys have been one of the prominent Accident Tolerant Fuel (ATF) cladding material candidates, primarily due to their excellent oxidation resistance in high-temperature steam conditions when compared to Zr-based alloys. Prototypic irradiation of fueled FeCrAl rods is a fundamental step in confirming the integral performance behavior during in-pile conditions. Early generation C26M cladding, fabricated through wrought metallurgy techniques, was fueled with UO2 fuel pellets and irradiated in a pressurized water loop in the Advanced Test Reactor (ATR) at the Idaho National Lab (INL) to a burnup (BU) of ∼25 GWd/tHM. After irradiation, the rodlets were nondestructively and destructively examined. Nondestructive examinations included visual exams, profilometry, and gamma scanning. These examinations highlight the unique deposits, BU profile of the rodlets as well as the migration path of fission gases within the rodlet, and typical diametrical morphology of fuel rodlets. During handling, brittle failure of one end cap on one pin occurred. Destructive examinations included microscopy and mechanical testing. Radial cross sections of the cladding were analyzed metallographically through light optical microscopy (LOM), scanning electron microscopy (SEM) highlighting a unique corrosion morphology and micro-cracking (∼20–30 μm past the main oxide layer) at the metal-oxide interface. Ring compression testing (RCT) was used to elucidate the mechanical property change of the FeCrAl cladding after neutron irradiation. In contrast to the non-irradiated material, which remained ductile at all test temperatures between 25 °C and 250 °C, the irradiated cladding fractured in a brittle manner at 50 °C and below. The results of the tests show that some challenges remain in the development of FeCrAl cladding for LWR cladding applications including improvement of in-reactor waterside corrosion performance and the retention of ductility after neutron irradiation.

Validation of MOOSE’s subchannel module using the AREVA FCTF heated bundle benchmark

The SubChannel Module, referred to as SCM henceforward in this document, is a subchannel code within the Multiphysics Object-Oriented Simulation Environment (MOOSE) developed at the Idaho National Laboratory (INL). SCM, is able to model flow through liquid-metal cooled, wire-wrapped fuel pin sub-assemblies, ordered in a triangular lattice. This work extends the existing validation done for these flow types and geometries and introduces validation for the newly implemented capability to model flows within a deformed duct. The subchannel duct deformation modeling approach consisted of reproducing the deformed duct effect on the geometric parameters of the boundary subchannels, i.e., modifying surface area, wetted perimeter and pin to duct gap. Then, the model was validated using experimental data taken from Areva’s fuel cooling test facility (FCTF).

Digital engineering implementation in nuclear demonstration and nonproliferation projects at Idaho National Laboratory

Digital engineering and digital twins are increasingly being used in nuclear energy projects with important impacts. At Idaho National Laboratory, these approaches have been applied in a variety of nuclear energy research, development, and demonstration projects, with key lessons and evolutions occurring for each. In this paper, we describe the use of digital engineering and digital twins in the Versatile Test Reactor design, National Reactor Innovation Center test beds, and nonproliferation analysis of the AGN-201 reactor design. We share key lessons learned for these projects related to tool selection, adoption and training, and working with existing assets versus beginning at the design phase. We also share highlights of future potential uses of digital twins and digital engineering, including using artificial intelligence to perform repetitive design tasks and digital twins to move towards semiautonomous nuclear power plant operations.

Protonic Ceramic Electrochemical Cell Manufacturing at Idaho National Laboratory

Protonic Ceramic Electrochemical Cells (PCECs), operating at intermediate temperatures of 400-600oC, show great potential for hydrogen production and versatile chemical manufacturing. As this technology rapidly evolves, scaling up production in both quantity and size is crucial. This step bridges the gap between fundamental academic research on small button cells (e.g., 10 mm to 1 inch) and higher-technology readiness level development of industry-standard units (> 10x10 cm2) for stack assembly. At Idaho National Laboratory, we've made significant strides in this transition by applying advanced manufacturing technologies and our expertise in materials science and engineering. Our approach integrates materials, structural, and interface engineering research and development into the cell fabrication process, guided by a stringent quality assurance and quality control (QA&QC) system. This ensures the reliable, repeatable, and durable performance of PCECs.

(Invited) Advancement in Electrosynthesis of Fuels and Chemicals Using H2O and/or CO2 and N2 at Intermediate Temperatures at Idaho National Laboratory

Fuels and commodity chemicals make up a significant share of the potential energy savings opportunities for decarbonized economy and society. New innovative technologies are needed that can reduce the amount of energy and carbon intensity currently required to produce the key fuels and chemicals or associated building blocks. Electrochemistry is playing an increasingly important role in manufacturing as we reduce energy use and address uncertain supply chains driven by industrial electrification. In comparison to processes driven by heat and pressure, electrochemically driven synthesis of fuels and chemicals provides unique opportunities for selective reactions and separations offering increased flexibility to manufacturing. As one of DOE’s premier multi-program energy research laboratories, Idaho National Laboratory’s (INL) mission is to ensure the nation’s energy security with safe, competitive and sustainable energy systems and unique national and homeland security capabilities. INL’s initiative, integrated energy systems (IES) provides carbon-free nuclear and renewable energy for downstream use in manufacturing and transportation. The manufacturing component of IES focuses on integrating clean, carbon-free electrons with waste/renewable heat and includes both manufacturing of electrochemical devices and implementing them in energy-efficient processes. In this talk, we will provide representative examples for the pillar of energy to molecules and materials (E2M2) by utilizing the protonic ceramic electrochemical cells/membrane reactors (PCEC/PCEMR). The electrochemical route using PCEC/PCEMR offers a solution of process intensification and low carbon intensity. We will also share our cost perspective when compared to other lower and higher temperature counterparts.

Challenges in Practical Button Cell Testing for Hydrogen Production from High Temperature Electrolysis of Water

High temperature electrolysis of water using solid oxide electrochemical cells (SOEC) is a promising technology for hydrogen production with high energy efficiency and may promote decarbonization when coupled with renewable energy sources and excess heat from nuclear reactors. Over the past several decades there have been extensive scientific and engineering studies on cell materials and degradation behaviors that have greatly improved current density, decreased total resistance, and lowered degradation rates. Although the technology is now at a near-commercial level, maintaining consistency in cell testing and minimizing variance in practical testing environments is an often overlooked but crucial topic. To promote high quality data collection, testing procedures and balance of plant component details are extremely important to consider. This work discusses some key factors affecting the reproducibility of practical SOEC testing on the button cell level, namely current collection layers, cell sealing procedures, the reliability of steam and hydrogen delivery systems, cell testing fixture design, and reduction procedures. To provide a baseline and a level of standardization for the SOEC community, this work also discloses details of the standard operating procedure and techniques adopted for o-SOEC testing at Idaho National Laboratory (INL).

Implementing Component Degradations into a Modelica Model of an iPWR System to Develop Health Monitoring Techniques

The economic operation of small modular reactors will partly rely on managing and reducing inspection and maintenance activities while supporting new operational paradigms like load-following. Turbine control valves throttle the steam from the steam generator into the steam turbine while maintaining the pressure within the steam generator at a constant set point. Degradation of these components could impact the ability to manage electrical power production. Utilizing the Idaho National Laboratory Hybrid repository and the Oak Ridge National Laboratory TRANSFORM library developed for multiphysics simulations in Dymola/Modelica, an integral pressurized water reactor system was modeled based on the available specifications of the NuScale power module. The effects of various component degradation modes have been implemented into the model in order to simulate faulted plant data during both steady-state and load-following operations. The fault modes resemble different physical fault modes that may occur at an operating nuclear power plant; a leaking turbine control valve and a valve actuator failure due to loss of hydraulic pressure have been implemented. A neural network autoencoder is employed in conjunction with statistical analysis, namely, simple signal thresholding (SST) or sequential probability ratio testing (SPRT), to identify the presence of a fault. Fuzzy logic is additionally employed in a novel and promising manner to classify the state of the system based on the cumulative sum of the neural network residuals. SST and SPRT are both successfully validated using healthy data and proved capable of identifying both fault types; fuzzy logic identified the false positives and classified the faulted data correctly.

X-Ray Computed Tomography at Idaho National Laboratory’s Irradiated Materials Characterization Laboratory

X-Ray Computed Tomography at Idaho National Laboratory’s Irradiated Materials Characterization Laboratory

Thermal Model of the AGR-5/6/7 Experiment with Offset Gas Gaps

Because of their high impact on calculated temperatures, the offset gas gaps between the graphite holder and capsule wall are evaluated for the thermal model of the Advanced Gas Reactor 5/6/7 (AGR-5/6/7) irradiation experiment. Fuel compact temperatures are a crucial factor in assessing the irradiation performance of tristructural isotropic (TRISO)-coated-particle fuel for use in high-temperature gas-cooled reactors. TRISO fuel irradiation performance is a core tenet of the U.S. Department of Energy’s Advanced Gas Reactor Fuel Development and Qualification Program, conducted in the Advanced Test Reactor at Idaho National Laboratory. Specifically, the irradiation experiments were intended to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support the development of fuel performance models and codes, and provide irradiated fuel and materials for postirradiation examination and safety testing. The thermal model used to predict the fuel compact temperatures was a three-dimensional finite element model that was subject to uncertainty. The most dominant factor in the uncertainty pertaining to calculated fuel temperatures is the gas gap uncertainty, which is caused by the graphite holder nub-to-capsule wall clearance. In AGR-5/6/7 capsules, the graphite holders can be shifted away from being perfectly centered inside the stainless steel capsule walls. This is due to irradiation-induced graphite shrinkage or to insufficient nubs on the graphite holder leading to azimuthally varying gas gaps during irradiation. A method was developed to find the best offset magnitude and direction at the top and bottom of the graphite holders in Capsules 1 and 2, which experienced higher temperatures than originally predicted among the five AGR-5/6/7 axially dispersed capsules irradiated between 2018 and 2020. The differences between the calculated and the measured temperatures of thermocouples (TCs) placed in the graphite holder near the fuel compacts were used as the basis for the holder offset determination. The best-fit offset assumes that the root-mean-square error between the measured and calculated TC temperatures is a minimum value. Typical offsets in the range of 0.002 to 0.004 in. increase the fuel compact temperatures in the range of 100°C to 200°C when compared to the zero offset for Capsules 1 and 2. Additionally, because of irradiation-induced changes in the capsule geometries, the holder offsets change throughout the irradiation period.

Design Basis Model for Hosting Small Modular Reactors

AbstractTo provide energy security and head off further increases in global temperatures, an aggressive transition from fossil fuels to other types of energy implies the need to possibly construct hundreds of nuclear power plants in the near future. However, the real and perceived risks of nuclear energy remain a significant impediment to that transition.This paper describes a comprehensive work process that combines the rigor of model‐based systems engineering (MBSE) with 1) the Idaho National Laboratory's (INL) decades of experience with small reactors and with 2) modern project delivery processes. The objective is to reduce the risks of building new facilities or converting existing facilities to nuclear power generation.

Mission Results: Creating a 3D Map of a Very High Radiation Confined Space Using the LiDAR-Equipped Elios 3 Drone

A prototype configuration of an Elios 3 indoor inspection drone, made by Flyability, comprising a lightweight light detection and ranging (LiDAR) system and a wide-range, electronic dosimeter was developed and tested to quickly measure radiation levels and collect three-dimensional (3D) spatial data from within a very high radiation nuclear waste storage facility at the U.S. Department of Energy−operated Idaho National Laboratory (INL) site. The compact drone configuration was used for inspecting and collecting data from areas difficult or hazardous to access using conventional methods. Validation testing of the prototype drone configuration was performed, including maneuverability, radiation tolerance, dosimeter integration, and elevated ambient temperature testing. The drone configuration was shown to operate successfully in an ionizing radiation field as high as 100 Gy/h (10 000 rad/h), with an accumulated radiation dose of 40 Gy (4000 rad) and ambient air temperatures of 54°C (130°F). On November 16, 2022, the LiDAR and dosimeter-equipped prototype drone was used to successfully collect 3D spatial data and radiation measurements from within a very high radiation nuclear waste storage facility at the INL site, where it encountered radiation levels up to 7 Gy/h (700 rad/h). The mission, which included three flights of the prototype drone configuration, represents the first-ever drone flight within a very high radiation nuclear waste storage facility. On the third flight into the storage facility, the drone was unable to produce adequate lift necessary to fly back out of the facility. The lessons learned suggest the loss of adequate lift was due to the added weight of the dosimeter payload and the increasing temperature of the air inside the facility due to radioactive decay. Although the prototype drone was not retrievable, the drone’s wireless transmission system successfully transmitted the radiation measurements to engineers outside of the facility.

A new approach to monitoring solvent extraction processes for the nuclear industry

As part of an initiative to steward research, development, and innovation into the nuclear fuel cycle, Idaho National Laboratory is building the Beartooth test bed. Beartooth will include a cascade of centrifugal contactors, glove box lines, and solidification and dissolution equipment to aid in the progression of novel separation techniques and to provide hands-on opportunities to early-career separation engineers. Beartooth will incorporate novel monitoring techniques using sensors and machine learning algorithms to inform process operators of separation conditions. This research is examining unconventional monitoring technologies such as incorporating acoustic microphones, vibration sensors, infrared cameras, red-green-blue color sensors, among others into nuclear separation processes. Research is ongoing in the use of machine learning methods to detect faults and alert operators of typical and anomalous events. Results from this research have the potential to impact safeguards-by-design efforts and real-time decision making. This overview will detail preliminary measurement results from acoustic, vibration, and color sensors.

Uranium-zirconium hydride nuclear fuel performance in the NaK-cooled MARVEL microreactor

The Microreactor Applications Research Validation and EvaLuation (MARVEL) Project is producing a high temperature NaK-cooled nuclear reactor test bed at the Idaho National Laboratory to ultimately improve integration of microreactors to end-user applications. This ambitious effort seeks to design, authorize, construct, test, and operate the reactor within five years. The computational software package chosen for high fidelity fuel performance modeling, BISON, has been upgraded to model the performance of the MARVEL reactor's selected fuel system: 304 stainless steel-clad U-ZrHx. In this manuscript, we begin by describing the MARVEL fuel element. The known properties and relationships of the fuel element constituents are recapitulated from data available in the literature. These historical relationships assume a particular U-ZrH microstructure which previously was not well-understood; therefore, the fuel's microstructure is experimentally confirmed using electron microscopy and X-ray diffraction. This information is then used to analyze the thermomechanical performance of the MARVEL fuel element during a hypothetical extreme accident scenario (a loss of flow accident during a loss of coolant pressure accident) using the nuclear fuel performance code BISON. The analysis shows that the primary phenomenon which could compromise the structural integrity of the MARVEL fuel element is the same as for TRIGA reactors: high temperature internal gas overpressurization-induced rupture of the cladding arising from excessive hoop stresses. Despite modeling the most severe MARVEL accident scenario, the highest anticipated temperature in the fuel, about 718 °C, results in hoop stresses of less than 10 MPa. Based on this analysis, a conservative steady state fuel meat temperature limit of 900 °C precludes cladding damage. The newly incorporated fuel performance simulation capabilities developed in this work may also be extended to model this fuel system under other conditions or reactors.