Idaho National Laboratory Research Articles

Design and testing of an enriched uranium fueled molten salt irradiation vehicle

Molten salt reactors (MSRs) have garnered increasing attention recently with several demonstration efforts on the way. A key challenge to the licensing basis for these reactors is the lack of experimental data on fueled salts. This is expected to be crucial to the safety evaluation and licensing basis of reactors of this type deployed in the future. While capability for irradiating molten salts has been reestablished in the recent decade, no enriched fuel irradiation capability has been developed and tested as of yet. A new experiment vehicle under development at Idaho National Laboratory (INL) is presented here. The Molten-salt Research Temperature-controlled Irradiation (MRTI) experiment was developed to host enriched uranium bearing salt samples to be irradiated at a test reactor within the lab complex. One of the key scientific objectives is to provide irradiated salt samples for post-irradiation examination (PIE) to study the impact of fission product generation and neutron/gamma radioactivity on the salt solution and salt-facing wall material. This paper provides a detailed overview of the mechanical design of the experiment, followed by an overview of the fabrication and assembly of an initial prototype vehicle (with non-fuel-bearing salt). A summary of the key analyses conducted as a part of the performance and safety evaluation is then provided. Lastly, an overview of the test conducted in prototypic out-of-pile (non-neutron) environment are shown. These evaluations provide the foundation for a planned irradiation of an enriched uranium-bearing chloride salt sample in the near term. The upcoming irradiation will contain 13 cm3 of UCl3–NaCl salt (93% enrichment) generating around 20 W/cm3 of fission energy during irradiation and a temperature range that can be contained between bounds of 525–900 °C.

A Review of Candidates for a Validation Data Set for High-Assay Low-Enrichment Uranium Fuels

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.

Engineering-Scale Integrated Energy System Data Projection Demonstration via the Dynamic Energy Transport and Integration Laboratory

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.

A non-fueled nuclear-heated rod for in-pile transient boiling studies

Separate-effects boiling experiments have recently been conducted in the Transient Reactor Test Facility at Idaho National Laboratory to investigate transient heating and irradiation effects on cladding-to-coolant heat transfer. Specifically, transient critical heat flux (CHF) remains an important area of research, and better understanding of this phenomenon has potential for improving predictive models related to operational and safety limits. Consequently, this knowledge is expected to improve efficiency of light-water reactor operations. A novel borated nuclear-heated rodlet (BNHR) was designed to enable observation of transient cladding-to-coolant heat transfer phenomena. The final BNHR design takes a surrogate approach, wherein nuclear heating is induced by 10B(n, α) reactions rather than derived from fissions in a fueled specimen. The structure of the BNHR consists of a hollowed out borated (Bnat ∼ 2.05 wt%) stainless steel tube with an hourglass-shaped outer surface, capped at both ends with non-borated stainless steel. This geometry allows for inner-rodlet instrumentation and generation of the highest nuclear heating rates near the center of the rodlet to ensure onset of boiling near instrumentation for real-time observation. A novel approach to measuring the nuclear energy deposition rate in the BNHR separate and apart from the influence of the coolant, termed the n-α thermometer, is also detailed in this paper. This device has demonstrated excellent repeatability, and measurements indicate predictive modeling results for energy deposition in the BNHR rod agree within a 10% margin of the experiment measurements. These results give confidence that the BNHR design has successfully met experiment objectives.

Post-irradiation examination of the Sirius-1 nuclear thermal propulsion fuel test

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.

Current status of MELCOR 2.2 for fusion safety analyses

MELCOR is an integral code developed by Sandia National Laboratories (SNL) for the US Nuclear Regulatory Commission (USNRC) to perform severe accident analyses of Light Water Reactors (LWR). More recently, MELCOR capabilities are being extended also to analyze non-LWR fission technologies. Within the European MELCOR User Group (EMUG), organized in the framework of USNRC Cooperative Severe Accident Research Program (CSARP), an activity on the evaluation of the applicability of MELCOR 2.2 for fusion safety analyses has been launched and it has been coordinated by ENEA. The aim of the activity was to identify the physical models to be possibly implemented in MELCOR 2.2 necessary for fusion safety analyses, and to check if those models are already available in MELCOR 1.8.6 for fusion version, developed by Idaho National Laboratory (INL). From this activity, a list of modeling needs emerged from the safety analyses of fusion-related installations have been identified and described. Then, the importance of the various needs, intended as the priority for model implementation in the MELCOR 2.2 code, has been evaluated according to the technical expert judgement of the authors. In the present paper, the identified modeling needs are discussed. The ultimate goal would be to propose to have a single integrated MELCOR 2.2 code release capable to cover both fission and fusion applications.

The advanced characterization, post-irradiation examination, and materials informatics for the development of ultra high-burnup annular U-10Zr metallic fuel

U-Zr metallic fuel is a promising fuel candidate for Gen Ⅳ fast spectrum reactors. Previous experimental irradiation campaigns showed that the sodium thermal bonded U-10Zr fuel design can achieve a burnup of 10% fissions per initial heavy metal atom (FIMA). Advanced metallic fuel designs are pushing the burnup limit to 20% or even 30% FIMA. To achieve the higher burnup and eliminate the pyrophoric sodium, a prototypical annular fuel has been designed, fabricated, clad with HT-9 in the Materials and Fuels Complex, and irradiated in the Advanced Test Reactors of Idaho National Laboratory (INL) to a peak burnup of 3.3% FIMA. During irradiation, the mechanical contact between fuel and cladding acts as a thermal bond. The irradiation lasted for 132 days in the reactor. Recently, the archived fresh and irradiated fuel samples were characterized using advanced characterization capabilities in the Irradiated Materials Characterization Laboratory (IMCL) of INL. This article summarizes the results of advanced characterization and computer vision-based materials informatics to reveal the irradiation effects on U-Zr metallic fuel. Future work will focus on further implementation of advanced characterization and statistical data mining to improve the fidelity of fuel performance modeling and support U-Zr metallic fuel qualification for fast spectrum reactors.

Weak State as a Security Threat: Study Case of El Salvador (2014-2019)

Weak State as a Security Threat: Study Case of El Salvador (2014-2019)

Total Energy for the SOFC and SOEC

The Solid Oxide Fuel Cell (SOFC) Program at the National Energy Technology Laboratory (NETL) managed by the U.S. Department of Energy (DOE) Office of Fossil Energy and Carbon Management (FECM) is currently developing low-cost SOFC and Solid Oxide Electrolysis Cell (SOEC) systems. This paper develops the Total Energy (TE) (kilowatt-hours per kilogram hydrogen, kWh/kgH2 ) for the SOEC and SOFC. The Total Energy includes heat input, exergetic flows, enthalpy of vaporization, pressurization, heat loss, area specific resistance, etc. The SOEC Total Energy developed at NETL, as it would happen, correlates well with the Idaho National Laboratory (INL) proven SOEC performance of forty-five kilowatt-hours per kilogram hydrogen at twenty bars, 1.3 volt and 725oC. Total Energy is necessary for designing, predicting, and planning for SOEC and SOFC performance and cost.

High-temperature irradiation-resistant thermocouple instability model for in-pile reactor use

This article presents an instability model for the high-temperature irradiation-resistant thermocouple (HTIR-TC). Here the term instability defines the superposition of both drift and inhomogeneity of TC thermoelements occurring simultaneously. The HTIR-TC is an advanced thermocouple (TC) that uses the refractory metals niobium and molybdenum as sensing thermoelements for generating electromotive force (EMF) in a field of neutrons and at temperatures upward of 1,600°C. In the Advanced Gas Reactor (AGR) 5/6/7 tests conducted at Idaho National Laboratory’s Advanced Test Reactor (ATR), the HTIR-TCs showed low to moderate instability throughout the life of the test. The instability model reveals that HTIR-TCs can, when the operating temperature of the reactor fuel is normal, maintain performance throughout an 18-month refueling cycle typical of nuclear power plants, reflecting an instability of less than ±1%. The HTIR-TC is also qualified for incorporation into a test fixture during the testing of new fuels.

MELCOR-TMAP: The integration of MELCOR for fusion and TMAP4 for fusion reactor systems safety analysis and tritium inventory tracking

The design of fusion reactors requires the prediction of radioactive tritium migration during normal operating conditions and accident scenarios. Safety standards such as those from the U.S. Department of Energy (DOE) require that fusion reactor designs limit the release of hazardous material to the environment. In addition to safety requirements, the rarity of tritium supplies for burning D-T fusion plasma requires tritium breeding blankets. Computational modeling of migration of tritium and various reaction products throughout the fusion reactor serves to create designs that fulfill these requirements. MELCOR-TMAP, owned and developed at Idaho National Laboratory (INL), is an integrated system-level engineering code adapted for the analysis of fusion reactor systems. The history of the conception and evolution of the MELCOR-TMAP code, as well as its current capabilities, recent updates including a corrected composite material wall nodalization scheme, and plans for future development, are discussed.

Influence of microstructure and phase morphology on the stability of high temperature irradiation resistant thermocouples

Development of in-core instrumentation is driven by the pursuit of safer and more economic energy production from both existing nuclear reactors and Generation IV reactor designs. Idaho National Laboratory (INL) has developed high temperature irradiation resistant thermocouples (HTIR-TCs) for temperature sensing inside Generation IV nuclear reactors. These thermocouples are composed of phosphorus-doped niobium (Nb-P) and lanthana-doped molybdenum (Mo-LaO) thermoelements, an alumina (Al2O3) insulation, and a niobium sheath. HTIR-TCs require an initial heat treatment exceeding the maximum service temperature to stabilize the generated electromotive force (EMF) signal; however, the mechanism behind this stabilization is not well understood. This work evaluates the impact of the stabilization heat treatment on the thermoelements' microstructures, chemical stability, and electrical properties to determine the mechanisms by which the EMF signal stabilization occurs. Accordingly, during the preliminary heat treatment, a secondary Nb3P phase formed within the Nb-P, along with an interaction region at the Al2O3/niobium interface. The formation of secondary phases within the niobium leg of the thermocouple causes an increase in the Seebeck coefficient. Stabilization of the HTIR-TC EMF signal was found to be dependent upon both the equilibrium of a diffusion interaction region at the Nb-P/Al2O3 interface and the formation of Nb3P precipitates.

Life Cycle Assessment of Innovative Carbon Dioxide Selective Membranes from Low Carbon Emission Sources: A Comparative Study.

Carbon capture has been an important topic of the twenty-first century because of the elevating carbon dioxide (CO2) levels in the atmosphere. CO2 in the atmosphere is above 420 parts per million (ppm) as of 2022, 70 ppm higher than 50 years ago. Carbon capture research and development has mostly been centered around higher concentration flue gas streams. For example, flue gas streams from steel and cement industries have been largely ignored due to lower associated CO2 concentrations and higher capture and processing costs. Capture technologies such as solvent-based, adsorption-based, cryogenic distillation, and pressure-swing adsorption are under research, but many suffer from higher costs and life cycle impacts. Membrane-based capture processes are considered cost-effective and environmentally friendly alternatives. Over the past three decades, our research group at Idaho National Laboratory has led the development of several polyphosphazene polymer chemistries and has demonstrated their selectivity for CO2 over nitrogen (N2). Poly[bis((2-methoxyethoxy)ethoxy)phosphazene] (MEEP) has shown the highest selectivity. A comprehensive life cycle assessment (LCA) was performed to determine the life cycle feasibility of the MEEP polymer material compared to other CO2-selective membranes and separation processes. The MEEP-based membrane processes emit at least 42% less equivalent CO2 than Pebax-based membrane processes. Similarly, MEEP-based membrane processes produce 34-72% less CO2 than conventional separation processes. In all studied categories, MEEP-based membranes report lower emissions than Pebax-based membranes and conventional separation processes.

Optimization of Conductive Fins to Minimize UO2 Fuel Temperature and Radial Temperature Gradient

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.

Bukelismo. Regres demokracji w Salwadorze w okresie rządów Nayiba Bukele

BUKELISMO: THE REGRESSION OF DEMOCRACY IN EL SALVADOR DURING THE TERM OF NAYIB BUKELE The aim of the article is to analyze the process of autocratization of the El Salvador’s political system after 2019, as well as to show the specifics of the government of Nayib Bukele. The author, based on the systemic and institutional‑legal methods, and to a lesser extent empirical and comparative methods, verifies two research hypotheses, according to which (1) there was a violent and radical regression of the political system and a permanent transition from democracy to autocratic rule in El Salvador during the term of Nayib Bukele; (2) bukelismo is a classic combination of populism, militarism, and authoritarian behavior of the leader. The Salvadoran case of democratic regression has been presented in the broader context of anti‑democratic changes in the Latin American region. In the article, the author focuses on several key problems: systemic changes resulting in an imbalance and the separation of powers; mechanisms for controlling the executive power; populism and control of the information space by the president; and the militarization of politics.

Asymptotic hydrogen redistribution analysis in yttrium-hydride-moderated heat-pipe-cooled microreactors using DireWolf

Yttrium hydride (YHx) is one of the materials being considered for moderating thermal and epithermal nuclear microreactors. One potential issue with YHx use is that the hydrogen redistributes in the hydride when thermal and concentration gradients are present. This hydrogen redistribution leads to spatial gradients in the hydrogen concentration, thus affecting neutron transport in the reactor. By building upon observations in prior works, this paper aims to gain a better understanding of the reactivity feedback associated with such hydrogen redistributions. In particular, we wish to understand the sign (+/−) of the hydrogen redistribution neutronic feedback, its order of magnitude, and its underlying physical causes. To achieve this goal, the DireWolf multiphysics software driver was used to solve the coupled radiation transport, heat transfer, heat pipe two-phase flow, and hydrogen redistribution equations for the Simplified Microreactor Benchmark Assessment (SiMBA) problem, a full-core microreactor numerical benchmark developed at Idaho National Laboratory.

Quantitative evaluation of common cause failures in high safety-significant safety-related digital instrumentation and control systems in nuclear power plants

• A comprehensive quantitative evaluation of common cause failures is performed for safety-critical digital control systems; • Both hardware and software common cause failures are tracked and identified; • Failure probabilities of digital control systems are estimated via a multiscale reliability analysis approach. • The impact of common cause failures to plant safety is evaluated. Digital instrumentation and control (DI&C) systems at nuclear power plants (NPPs) have many advantages over analog systems. They are proven to be more reliable, cheaper, and easier to maintain given obsolescence of analog components. However, they also pose new engineering and technical challenges, such as possibility of common cause failures (CCFs) unique to digital systems. This paper proposes a Platform for Risk Assessment of DI&C (PRADIC) that is developed by Idaho National Laboratory (INL). A methodology for evaluation of software CCFs in high safety-significant safety-related DI&C systems of NPPs was developed as part of the framework. The framework integrates three stages of a typical risk assessment—qualitative hazard analysis and quantitative reliability and consequence analyses. The quantified risks compared with respective acceptance criteria provide valuable insights for system architecture alternatives allowing design optimization in terms of risk reduction and cost savings. A comprehensive case study performed to demonstrate the framework's capabilities is documented in this paper. Results show that the PRADIC is a powerful tool capable to identify potential digital-based CCFs, estimate their probabilities, and evaluate their impacts on system and plant safety.

EBR-II MOX Fuel Characterization Enabling ARES Phase I Testing

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.

Automation levels for nuclear reactor operations: A revised perspective

This work serves to propose updated levels of automation for nuclear reactor operations, as a result of considering long-term economic and commercial ambitions of the advanced reactor developer community. As in other fields such as road-going vehicles and aviation, reactor technologies can benefit from modern automation through the resulting reduction in operations and maintenance costs, while still maintaining the current industry standards regarding safety, resilience, reliability, overall performance, and the capacity for root-cause analysis. The current guidelines on automation levels, as published by the U.S. Nuclear Regulatory Commission in Section 9 of NUREG-0700, reflect outdated design principles that implicitly limit the potential of automation innovation for reactor operations, particularly in regard to advanced reactors intended to operate in remote locations or be used for off-grid applications. Motivated by the operational paradigms anticipated for future reactor designs, we propose a six-level approach that aligns with contemporary automation concepts as well as automation level definitions from other non-nuclear safety–critical industries. These levels build upon the current guidelines in order to enable next-generation nuclear reactor technologies to become increasingly economically competitive and commercially viable relative to competing power generation sources. Using a hypothetical heat removal reactor transient, we provide examples of how the human–machine interactions change at each level of automation, ranging from traditional operator control (Level 0) to operator-free unattended operations (Level 5)—the latter being one of the key attributes proposed at the Fission Battery Initiative led by Idaho National Laboratory. Finally, we critically examine the identified challenges, knowledge gaps, and enabling technologies to achieve advanced levels of automation.

The Virtual Test Bed (VTB) Repository: A Library of Reference Reactor Models Using NEAMS Tools

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.