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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. [ABSTRACT FROM AUTHOR]

Bromine-82, Potassium-42 and Copper-64 have been successfully adopted as radioactive surrogates for outdoor large area contamination training. The goal of this project was to discover new materials that could supplement potassium bromide (KBr) and copper pellets in radiological dispersal device (RDD) training events to reduce the down time of the training fields and to broaden the toolbox of the RDD surrogates at Idaho National Laboratory. Of the ten different materials investigated, sodium nitrite, gallium metal, and gallium oxide presented the greatest promise due to their ability to meet or exceed selection criteria: half-lives (between 10 and 36 h), limited impurities (99.9%), and favorable National Fire Protection Agency (NFPA) rating. [ABSTRACT FROM AUTHOR]

Nuclear Thermal Propulsion (NTP) systems hold promise in reducing transit times for exploration of the solar system by both crewed and uncrewed missions. NTP systems currently under investigation include a once-through high-temperature gas-cooled fission reactor to provide thermal energy to heat the coolant, which also serves as the propellant. The fuel system consists of angular UN fuel particles dispersed in a matrix of W/Re, creating a ceramic and metallic composite or cermet, which has been irradiated in Idaho National Laboratory's (INL) Transient Reactor Test Facility (TREAT). This enabled evaluation of these materials under representative nuclear heating rates (∼95 K/s) and peak surface temperatures (∼2527 K). These surface conditions were indicative of achievement of the target peak internal temperature of approximately 2850 K. These tests named Sirius-1 (UN–W/Re), have been irradiated and this paper will present post-irradiation examination results. The Sirius-1 test produced minor, stable cracks in the fuel specimen and spalling of surface material. Volatilized uranium 'soot' was found deposited on the inner wall of the irradiation capsule indicating loss of some fissile material from the fuel specimen. Spalling from the surfaces was also noted upon visual inspection. Uranium diffusion from the fuel particles resulted in the formation of U/Re phases and the production of a laminar microstructure on the edge of the fuel system. Overall, the specimen was stable and remained within a coolable geometry upon conclusion of multiple thermal cycles to prototypical operating temperatures. • A cermet nuclear fuel of UN particles in a W/Re matrix was irradiated and examined. • The sample material is a candidate for a nuclear thermal propulsion system. • The irradiation used prototypic heating rates and peak temperatures. • Post examination revealed that the fuel specimen remained largely intact. • Some localized zones of intermetallic W–Re sigma phase existed. [ABSTRACT FROM AUTHOR]

The interpretation of the results of hydrogeological field observations and the modeling of fractured porous subsurface media is often conducted using dual-porosity and/or dual-permeability concepts. These concepts, however, do not consider the effects of spatial and temporal variations and uncertainties, or fuzziness, in the evaluation of the subsurface flow characteristics of fractured porous media. The goal of the paper is to introduce a concept of fuzzy dual permeability of fractured porous media based on the fuzzy system analysis of the results of ponded infiltration tests in fractured basalt. The author revisited the results of the tests conducted in areas close to the Idaho National Laboratory (INL), Idaho, USA: small-scale (approximately 0.5 m2) ponded tests at the Hell's Half Acre site, mesoscale (56 m2) ponded tests at the Box Canyon site, and a large-scale infiltration test (31,416 m2) at the Radioactive Waste Management Complex at INL. Methods of fuzzy clustering and fuzzy regression were applied to describe the time-depth waterfront penetration and to characterize the phenomena of rapid flow through a predominantly fractured component and slow flow through a predominantly porous matrix component. The concept of fuzzy dual permeability is presented using a series of fuzzy membership functions of the waterfront propagation with depth and time. To describe the time variation of the flux, a fuzzy Horton's model is presented. The developed concept can be used for the uncertainty quantification in flow and transport in geologic media. [ABSTRACT FROM AUTHOR]

Many advanced reactor concept designs rely on high-assay low-enriched uranium (HALEU) fuel, enriched up to approximately 19.75% 235U by weight. Efforts are underway by the US government to increase HALEU production in the United States to meet anticipated needs. However, very few data exist for validation of computational models that include HALEU, beyond a few fresh fuel benchmark specifications in the International Reactor Physics Experiment Evaluation Project. Nevertheless, there are other data with potential value available for developing into quality benchmarks for use in data- and software-validation efforts. This paper reviews the available evaluated HALEU fuel benchmarks and some of the potentially relevant benchmarks for fresh highly enriched uranium. It then introduces experimental data for HALEU fuel irradiated at Idaho National Laboratory, from relatively recent irradiation programs at the Advanced Test Reactor. Such data should be evaluated and, if valuable, collected into detailed benchmark specifications to meet the needs of HALEU-based reactor designers. [ABSTRACT FROM AUTHOR]

The objective of this study is to demonstrate and validate the Dynamic Energy Transport and Integration Laboratory (DETAIL) preliminary scaling analysis using Modelica language system-code Dymola. The DETAIL preliminary scaling analysis includes a multisystem integral scaling package between thermal-storage and hydrogen-electrolysis systems. To construct the system of scaled equations, dynamical system scaling (DSS) was applied to all governing laws and closure relations associated with the selected integral system. The existing Dymola thermal-energy distribution system (TEDS) facility and high-temperature steam electrolysis (HTSE) facility models in the Idaho National Laboratory HYBRID repository were used to simulate a test case and a corresponding scaled case for integrated system HYBRID demonstration and validation. The DSS projected data based on the test-case simulations and determined scaling ratios were generated and compared with scaled case simulations. The preliminary scaling analysis performance was evaluated, and scaling distortions were investigated based on data magnitude, sequence, and similarity. The results indicated a necessity to change the normalization method for thermal storage generating optimal operating conditions of 261 kW power and mass flow rate of 6.42 kg/s and the possibility of reselecting governing laws for hydrogen electrolysis to improve scaling predictive properties. To enhance system-scaling similarity for TEDS and HTSE, the requirement for scaling validation via physical-facility demonstration was identified. [ABSTRACT FROM AUTHOR]

Pretransient characterization was performed for the Experimental Breeder Reactor II (EBR-II) mixed-oxide (MOX) fuel pellets from the SPA-2/-2B Operational Reliability Testing collaboration between Japan and the United States. Continued collaboration under the Advanced Reactor Experiments for Sodium Fast Reactor Fuels project will investigate the transient performance of these rods in the Transient Reactor Test facility at Idaho National Laboratory in the MOXTOP-THOR experiment. The results will fill a gap in existing transient performance data for MOX as these rods have a peak burnup of 14.3 at. % (~134.4 GWd/t) in the EBR-II. Fuel pellet properties were gathered from available resources and their irradiation and decay history evaluated. Further reactor physics calculations were performed to support the experiment design, reactor operations, and safety analyses necessary to enable the programmatic success of this effort. Of the three irradiated fuel pins, two will undergo transient testing, and all three will undergo post-irradiation examination. The methodology development and analysis activities utilized in this paper enable current experiment design work and provide the pathway through which measured data of this type can be further evaluated. [ABSTRACT FROM AUTHOR]

With the next generation of nuclear reactors under development, modeling and simulation tools are being developed by the U.S. Department of Energy to support their design, licensing, and future operation. Mirroring the physical test beds currently under construction (i.e., Demonstration and Operation of Microreactor Experiments, known as DOME, and Laboratory for Operating and Testing in the United States, known as LOTUS), the Virtual Test Bed was launched by the National Reactor Innovation Center in collaboration with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program to support the advanced reactor community. This collaborative effort, which involves multiple teams at both Idaho National Laboratory and Argonne National Laboratory, aims to use state-of-the-art simulation tools to model a wide range of reactor designs. These models are automatically tested to ensure their continued functionality as the tools are further developed. Examples are extensively documented, each acting as a tutorial for applying the relevant NEAMS tools to that reactor design. Currently, five advanced reactor types (with a total of 12 specific design subvariants) are simulated by a variety of models. These models range from steady-state, core multiphysics simulations to integrated plant analysis during loss-of-flow transients. To our knowledge, this is the first publicly available library of multiphysics advanced reactor models distributed with extensive documentation and maintained through continuous integration. [ABSTRACT FROM AUTHOR]

This paper describes ongoing work within the Light Water Reactor Sustainability pathway at Idaho National Laboratory (INL) to optimize the security and cost of nuclear power plants. It introduces the dynamic risk assessment tool developed at INL, Event Modeling Risk Assessment using Linked Diagrams (EMRALD). EMRALD is leveraged to optimize the security posture of a nuclear power plant by integrating force-on-force (FOF) simulations and operator mitigation actions, including dynamic and flexible coping strategies (FLEX). To illustrate the methodology, four attack scenarios are modeled in a commercially available FOF simulation tool using a hypothetical nuclear power plant facility. The simulation results provide valuable insights into possible attack outcomes, as well as the probabilistic risk of a core damage event given these outcomes. Safety mitigation procedures are modeled in EMRALD dependent on the attack outcomes by considering human operator uncertainties. The results demonstrate that the number of armed responders can be optimized, while still maintaining the same protection level as the initial security posture. The proposed modeling and simulation framework of integrating FLEX equipment with FOF models enables the nuclear power plants to credit FLEX portable equipment in the plant security posture, resulting in an efficient and optimized physical security system. [ABSTRACT FROM AUTHOR]

To further the development of low-enriched uranium fuels, precedence has been placed on delivering the same amount of power while lowering the fuel temperature and radial temperature gradient. To address this, modeling efforts have resulted in a novel design featuring conductive fins of varying thermal conductivities and geometries inserted into the fuel matrix. These conductive inserts were not allowed to exceed 6% of the original fuel volume. This constraint was imposed due to other designs displacing 10% of fuel volume. A parametric study was performed that consisted of 2.56 million BISON simulations involving varying fin characteristics (i.e., fin thermal conductivity, number, and geometry) to determine the optimal geometric configuration for a desired amount of fuel volume displaced. The results from this study show that the thickness and length of each fin affect the fuel temperature and temperature gradient more than varying the number and thermal conductivity of the fins. The parametric study resulted in the development of an optimized combination to produce the lowest peak fuel temperature, lowest radial temperature gradient, and highest temperature reduction for the amount of original fuel volume displaced. The simulations presented in this work will eventually be compared with irradiation experiments of similar fuel designs at Idaho National Laboratory's Advanced Test Reactor. [ABSTRACT FROM AUTHOR]

Monica Regalbuto - Respect, sincerity, and concern for workers are essential With remarkably diverse experiences in a variety of positions in the Department of Energy (DOE) as well as the private sector, Dr. Monica Regalbuto has contributed to the development of innovative energy technologies as a researcher, leader, and manager in the field of nuclear energy. Dr. Monica Regalbuto leads the Nuclear Fuel Cycle Strategy, at the Department of Energy's (DOE) Idaho National Laboratory, the nation's leading center for nuclear energy research and development. Dr. Regalbuto served in the DOE's Office of Nuclear Energy becoming Deputy Assistant Secretary for the DOE's Nuclear Energy Fuel Cycle Technologies. [Extracted from the article]

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-ZrH x. 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. [ABSTRACT FROM AUTHOR]

After approximately 50 years, NASA is restarting efforts to develop nuclear thermal propulsion (NTP) for interplanetary missions. Building upon nuclear engine tests performed from the late 1950s to the early 1970s, the present research and testing focuses on advanced materials and fabrication methods. A number of transient tests have been performed to evaluate materials performance under high-temperature, high-flux conditions, with several more experiments in the pipeline for future testing. The measured data obtained from those tests are being used to validate the Griffin reactor multiphysics code for this particular type of application. Griffin was developed at Idaho National Laboratory (INL) using the MOOSE framework. This article describes the simulation results of the SIRIUS-CAL calibration experiment in the Transient Reactor Test Facility (TREAT). SIRIUS-CAL was the first transient test conducted on NASA fuels, and although the test was performed with a relatively low core peak power, the test specimen survived a temperature exceeding 900 K. Griffin simulations of the experiment successfully matched the reactor's power transient after calibrating the initial control rod position to match the initial reactor period. The thermal-hydraulics model largely matches the time-dependent response of a thermocouple located within the experiment specimen to within the uncertainty estimate. However, the uncertainty range is significant and must be reduced in the future. [ABSTRACT FROM AUTHOR]

In April 2018 four drums of depleted uranium sludge burst open at a US Department of Energy facility at Idaho National Laboratory. This echoed a previous incident in 2014, when a drum erupted with fire and spewed radionuclides at a nuclear waste repository in New Mexico. Such "drum breach" accidents have been characterized in official reports as "isolated events," and as "self‐initiated" and "spontaneous." Yet these descriptions misrepresent a temporal tension at the heart of their causation: during the Cold War, nuclear weapons production created latent socioecological hazards that, decades later, too often manifest during waste cleanup. This tends to result from 21st‐century scheduling pressures, fraught labor relationships, and neoliberal subcontracting arrangements. Thus, if we examine the temporal form taken by US nuclear technopolitics, we can recast drum breach accidents not as stand‐alone events, but as outcomes of systemic incentives to speed up waste‐cleanup projects beyond their organizational capacity without commensurately expanding their safety or oversight mechanisms. [nuclear waste, technopolitics, accidents, neoliberalism, temporality, security, Idaho, United States] [ABSTRACT FROM AUTHOR]

To evaluate radioactive tracer dilution and gamma spectroscopy as a safeguards technique for monitoring the mass of liquid salt in molten salt systems, the gamma spectroscopy data of salt samples from the pyroprocessing facilities in Idaho National Laboratory were acquired and analyzed. The primary focus is the uncertainty analysis of the gamma radioactivity of 154Eu isotope in the molten salts. The facilities include an electrorefiner for pyroprocessing spent oxide fuels (SOF-ER) and Mark-IV ER for processing the metallic fuels from Experimental Breeder Reactor-II (EBR-II). The relative gamma radioactivity uncertainty was consistently at 3% for 137Cs for both Mark-IV ER and SOF-ER salts. However, for 154Eu, it was 8% for SOF-ER salt and 7% for Mark-IV ER salt. The main reason for the higher uncertainty for 154Eu is believed to be the lower counting statistics in gamma spectroscopy. To improve the gamma radioactivity uncertainty of isotope 154Eu, a longer data acquisition time (12 h instead of 4 h) and a higher count rate gamma detector for gamma spectroscopy were tested and it was found that the uncertainty for 154Eu was significantly improved from 7% or 8–3%. [ABSTRACT FROM AUTHOR]

Digital engineering can fundamentally change the construction and operation of complex projects. Traditional engineering teams operate in siloed tools and disparate teams. Throughout a project lifecycle, data is translated manually or over brittle point‐to‐point connections, which increases the risk of silent errors. These potential errors can cascade across the effort, resulting in significant project delays and cost overruns. Furthermore, traditional engineering projects are not designed for the rapid digitization that allows for advanced analytic tools. Digital engineering can create a cohesive data "source of truth" to ensure a project is designed for the digital future. There is a need for a standard toolset to normalize and integrate project datasets. Idaho National Laboratory (INL) has developed DeepLynx, a crucial tool to mitigate this data management problem. This paper introduces INL's digital engineering approach and provides a concrete example of a digital engineering design effort currently taking place with Beartooth, a new nuclear material processing testbed project. This paper will provide the benefits, challenges, and lessons learned in using digital engineering on this complex system. The digital engineering approach is utilizing model‐based systems engineering, extended reality, and a digital twin. Finally, this paper will discuss the new possibilities provided by using a digital thread and twin during future operation of a nuclear fuel cycle testbed. [ABSTRACT FROM AUTHOR]

Perception towards nuclear energy is a vital factor determining the success or failure of nuclear projects. An online survey obtained attitudes toward nuclear energy, opinions on whether benefits of nuclear energy outweigh the risks, and views of using nuclear energy as an energy source. A total of 4318 participants from across the U.S. completed the survey. Logistic regression was used to predict perceptions of nuclear energy by participant demographics and geographical location. Participants living closest to Idaho National Laboratory (INL) were more likely to have positive attitudes towards nuclear energy (aOR: 7.18, p < 0.001), believe the benefits were greater than the risks (aOR: 4.90, p < 0.001), and have positive attitudes toward using nuclear energy as an electricity source (aOR: 5.70, p < 0.001), compared to people living farther from INL. Males and non-Hispanic white participants were more likely to have positive perceptions of nuclear energy. Developing and implementing awareness raising campaigns for people living further away from nuclear power plants, targeting females and Hispanic whites, may be key to improving the overall perceptions of nuclear energy. [ABSTRACT FROM AUTHOR]

Subchannel codes have been widely used for thermal-hydraulics analyses in nuclear reactors. This paper details the development of a novel subchannel code within the Idaho National Laboratory's (INL) Multi-physics Object Oriented Simulation Environment (MOOSE). MOOSE is a parallel computational framework targeted at the solution of systems of coupled, nonlinear partial differential equations, that often arise in the simulation of nuclear processes. As such, it includes codes/modules able to solve the multiple linear and nonlinear physics that describe a nuclear reactor, under normal operation conditions or accidents. This includes thermal-hydraulics, fuel performance, and neutronics codes, between others. A MOOSE-based subchannel code is a new addition to the fleet of INL-developed codes, based on the MOOSE framework. In this work, we present the derivation of the subchannel equations for a single-phase fluid, we proceed with the description of the algorithm that is used to solve these equations and describe how this algorithm was implemented within MOOSE. We also present how this code can be coupled to the BISON fuel performance code. Next, we verify the friction model and the turbulent mixing model. We calibrate the turbulent modeling parameters for momentum mixing and enthalpy mixing, C T , β . We validate the code using experimental results and last demonstrate the coupling capabilities using a simple example. [ABSTRACT FROM AUTHOR]

This study addresses some of the challenges in the modeling of solid oxide electrolysis cells (SOECs), particularly for feed gases containing low inlet steam-to-CO 2 concentration ratios. The common approach used for SOEC modeling is to neglect the CO 2 electrochemical reduction reaction. A comprehensive model validation versus experimental results presented by the Idaho National Laboratory (INL) is performed in this paper. Our validation results under various operating conditions show that the model deviation with experimental data increases above certain current densities, e.g., 1200 Am−2. It can also be seen that the electrochemical reaction involving CO 2 can only be neglected when the steam flow supplied to the cell is high enough to support the water-gas shift reaction. Suppose the concentration of inlet water supply is not enough to support the reverse water gas shift reaction. In that case, the electrochemical reduction of CO 2 has to be considered to avoid model-predictive results that are far from available experimental observations. • Solid oxide electrolysis cells modeling Challenges at low inlet steam-to-CO 2 concentration ratios are presented. • CO 2 Electrochemical reaction is negligible when the inlet steam flow is high to support the water-gas shift reaction. • Numerical results can be far from experimental depending on the fuel composition and inlet steam into a cell. [ABSTRACT FROM AUTHOR]

In today's nuclear industry, momentum towards the design, licensing, and construction of advanced nuclear demonstration plants, including fast reactors, is at a remarkably high level. However, there are currently no dedicated fast spectrum irradiation test facilities in the United States to support the development of fast spectrum technologies. As a result, a unique situation is developing where most of these plants will likely be designed by leveraging historic nuclear material technologies, but where the further optimization and advancement is impeded by the lack of fast neutron irradiation test facilities. While these circumstances present a challenge, there are some near-term opportunities that, if seized, can still help develop advanced fast reactor materials to a meaningful level of readiness to support future commercial fast reactors. In this paper, we assess the feasibility of using thermal neutron filtering materials in existing experiment positions in the Advanced Test Reactor (ATR) at Idaho National Laboratory and the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory to simulate fast reactor test environments for nonfuel test specimens. Items investigated include the incident neutron flux (both fast and thermal), the total neutron fluence and cumulative atom displacements, helium production rate due to thermal neutron capture in nickel, and the potential impact that the thermal neutron filter material has on the cycle length of a given reactor. It is concluded that while HFIR provides the highest fast flux of all the options investigated, it is limited in the amount of thermal neutron filtering material that can be introduced into an experiment position without significantly affecting the operation of the reactor. Irradiation in Outboard-A positions in the ATR was found to be the most realistic near-term experiment avenue due to having ample space for several capsules in a moderately fast flux. [ABSTRACT FROM AUTHOR]

• Tritium migration analysis program version 8 (TMAP8), the latest version of TMAP, was developed within the framework of the multiphysics object-oriented simulation environment (MOOSE), an open-source, dimension-agnostic, fully coupled, and fully implicit multiphysics platform featuring massively parallel computation capabilities. • Using TMAP8, tritium transport in a divertor monoblock was analyzed to elucidate the effects of pulsed operation (up to fifty 1600 s plasma discharge and cool-down cycles) on the tritium invessel inventory source term and ex-vessel release term (i.e., tritium retention and permeation) for safety analysis. • With its built-in message passing interface capability, TMAP8 can, in under 2 h, simulate tritium transport in three different layered materials (i.e., tungsten, copper, and copper-chromiumzirconium alloy) in 2D geometry, using a single device/computer with 10 cores. • This work is our initial effort toward developing a high-fidelity modsim capability for fusion energy systems. Tritium Migration Analysis Program version 8 (TMAP8), the latest version of TMAP, was developed within the framework of the Multiphysics Object-Oriented Simulation Environment (MOOSE). Created at Idaho National Laboratory (INL), MOOSE is an open-source, dimension-agnostic, fully coupled, and fully implicit multiphysics platform featuring massively parallel computation capabilities. Using TMAP8, tritium transport in a divertor monoblock was analyzed to elucidate the effects of pulsed operation (up to fifty 1,600 s plasma discharge and cool-down cycles) on the tritium in-vessel inventory source term and ex-vessel release term (i.e., tritium retention and permeation) for safety analysis. With its built-in Message Passing Interface capability, TMAP8 can, in under 2 h, simulate tritium transport in three different layered materials (i.e., tungsten, copper, and copper-chromium-zirconium alloy) in 2D geometry, using a single device/computer with 10 cores. The MOOSE-based TMAP8 code can leverage other MOOSE tools developed under the Nuclear Energy Advanced Modeling and Simulation program to perform tritium and thermal transport in complex geometries and multiphysics environments. And via its massively parallel computation, MOOSE will enable the fusion pilot plant designers to conduct high-fidelity multiphysics modeling for the design of the divertor and blanket systems as well as for the safety analysis. [ABSTRACT FROM AUTHOR]

Recently, microreactor designs have been receiving significant attention in the nuclear industry due to their potential advantages in certain applications. These nuclear reactor designs have been considered to provide reliable and sustainable power for on-site installation and operation. Microreactors may be utilized to provide heat and power to hydrogen production, remote communities, and industrial facilities such as military installations, disaster relief zones, and are being considered for underwater and deep space operation as well. However, these designs and concepts remain largely untested and unproven in the commercial industry. Further research and development are still required to prove microreactor designs are safe and reliable for commercial use. Different cooling technologies have been taken into consideration for microreactor concepts since the 1960s, mainly for federal space reactor projects such as LEGOLRCS, HOMER, and KRUSTY. This work provides thermal hydraulics and analysis for the Idaho National Laboratory's MAGNET (Micro-reactor Agile Non-nuclear Experimental Testbed) facility. The MAGNET facility is currently being developed to duplicate a microreactor design using heat pipe cooling technology. Our main goal is to examine the response of the test facility under steady-state and transient operation conditions. We constructed our own estimated model of the MAGNET geometry using a software coupling method made up of MOOSE/SAM software systems. The steady state results of this work have been published in a former article. The new results mainly focus on the transient states. By communicating with the Idaho National Laboratories, we upgraded the geometry of MAGNET heat pipes. This not only verifies the design of the facility under such conditions but also benchmarks the modeling capability of the MOOSE/SAM code system that can be potentially used to model other microreactor concepts in the future. • Heat pipe micro reactor could be energy source of hydrogen production. • MAGNET is currently the experimental heat pipe micro reactor facility. • Multi-physical platform MOOSE/SAM are used to conduct the calculation. • Transient cases include different heat pipe failure numbers and times. • The results explore the safety and reliability edges of the heat pipe micro reactor. [ABSTRACT FROM AUTHOR]

The Dragonfly mission to Titan represents the first use of a large-scale radioisotope power source in that environment; as such little is known of its potential effects on the surface or near-surface atmosphere. These effects represent backgrounds it will be necessary to subtract from the spacecraft's instrumentation data. Repurposing a Monte Carlo code model of the generator built for previous neutronic investigations with the addition of electron detectors allows for the determination of the degree to which the radiation fields from the generator may ionize the surrounding atmosphere and generate interference with the lander's conductivity measurements. Results showed good agreement with experimental data undertaken at the Idaho National Laboratory using the F-2 generator flown on the Perseverance Mars rover, and showed that the interference with readings taken in the daytime on Titan when atmospheric conductivity is expected to be low may be substantial. • MCNP Model of MMRTG, Dragonfly drone hull section & Titan/Earth atmospheres. • Ionization of dry Titan atmosphere greater than humid Earth atmosphere. • Conductivity within spacecraft hull significantly enhanced by backscatter. • Simulation results verified against empirical results taken at Idaho National Lab. • Open question of effect of methane humidity requires experimental work. [ABSTRACT FROM AUTHOR]