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At least 73 records · Page 4

MSR Proliferation Resistance and Physical Protection White Paper

Molten Salt Reactors (MSRs) have seen a resurgence of interest in the past decade around the world. Support for these activities is provided from both national and private sources. The largest difference from the 2011 GIF MSR PR&PP evaluation consequently is the transition from evaluating academic systems focused on exploring the technical potential of MSRs to those of companies and countries focusing on near-term deployment. A wide variety of designs currently exist ranging from solid to liquid-fueled designs, with salt processing on-site or off-site, and a variety of fuel choices. As such, the proliferation resistance and physical protection aspects will have significantly more variation depending on reactor design than the other advanced reactors. The rapid introduction and evolution of innovative MSR designs inevitably means that technology specific details of overview reports, such as this one, become rapidly outdated. Consequently, this report focuses on essential features required for any MSR rather than specific design aspects.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Viral Fate and Transport for COVID-19 - NVBL

The NVBL Viral Fate and Transport Team includes researchers from eleven DOE national laboratories and is utilizing unique experimental facilities combined with physics-based and data-driven modeling and simulation to study the transmission, transport, and fate of SARSCoV-2. The team was focused on understanding and ultimately predicting SARS-CoV-2 viability in varied environments with the goal of rapidly informing strategies that guide the nation’s resumption of normal activities. The primary goals of this project include prioritizing administrative and engineering controls that reduce the risk of SARS-CoV-2 transmission within an enclosed environment; identifying the chemical and physical properties that influence binding of SARS-CoV-2 to common surfaces; and understanding the contribution of environmental reservoirs and conditions on transmission and resurgence of SARS-CoV-2.

59 BASIC BIOLOGICAL SCIENCES↗

Viral Fate and Transport for COVID-19

The NVBL Viral Fate and Transport Team includes researchers from eleven DOE national laboratories and is utilizing unique experimental facilities combined with physics-based and data-driven modeling and simulation to study the transmission, transport, and fate of SARSCoV-2. The team was focused on understanding and ultimately predicting SARS-CoV-2 viability in varied environments with the goal of rapidly informing strategies that guide the nation’s resumption of normal activities. The primary goals of this project include prioritizing administrative and engineering controls that reduce the risk of SARS-CoV-2 transmission within an enclosed environment; identifying the chemical and physical properties that influence binding of SARS-CoV-2 to common surfaces; and understanding the contribution of environmental reservoirs and conditions on transmission and resurgence of SARS-CoV-2.

54 ENVIRONMENTAL SCIENCES↗

Comparative Analysis of Confidence Metrics for Nuclear Criticality Safety

Nuclear criticality safety standards provide guidance on the requirements and recommendations to establish confidence in computerized model results used to support operation with fissionable materials. By design, the guidance is not prescriptive, leaving the analysts free to determine how various sources of uncertainties are to be statistically aggregated. This report compares the analyses and key assumptions behind four notable methodologies documented in the nuclear criticality safety literature: the parametric, nonparametric, Whisper, and TSURFER methodologies. Because of the involved use of statistics entangled with heuristic recipes, the results of these methodologies are often difficult to interpret. Also, they are augmented by additional large administrative margins, eliminating the incentive to understand their differences. With the new resurgent wave of advanced nuclear systems focused on economizing operation—including advanced reactors, fuel cycles, and fuel concepts—there is a strong need to develop a clear understanding of uncertainties and their fusion methodologies to reduce uncertainties in a scientifically defensible manner. This report offers a deep dive into the various assumptions of the four noted methodologies, their adequacy, and their limitations, to provide guidance on developing confidence for the emergent nuclear systems. These systems are expected to be challenged by the scarcity of experimental data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Regulatory Treatment of Non-Core Sources of Radioactivity for Advanced Reactor Designs

The recent resurgence in advance (non-light water) reactor development has been paralleled by the development of risk-informed performance-based (RIPB) licensing pathways. Specifically, the creation of the RIPB Licensing Modernization Project (LMP) approach and subsequent endorsement by the U.S. Nuclear Regulatory Commission (NRC) now provides advanced reactor vendors with a defined RIPB method to develop an affirmative safety case for licensing. In addition, the Technology Inclusive Content of Applications Project (TICAP) has published guidance on developing a license application based on the LMP approach. To support the utilization of risk information as part of advanced reactor design and licensing efforts, the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) Joint Committee on Nuclear Risk Management (JCNRM) has developed a probabilistic risk assessment (PRA) standard for advanced reactors. The standard, which was formerly approved by the American National Standards Institute (ANSI) in 2021 and recently endorsed by the NRC in trial use Regulatory Guide (RG) 1.247, is an integral standard, covering from initiating events to offsite consequence. A major feature of the standard is that it permits the inclusion of any source of radioactivity material at the site within the plant PRA. Therefore, non-core sources of radioactivity, such as fuel storage, fuel processing, and purification systems, can be included within a single comprehensive plant PRA. For those advanced reactor vendors utilizing a RIPB licensing approach, there is an opportunity to include the non-core sources of radioactivity within the RIPB framework for licensing decision-making, such as the categorization of events, classification of structures, systems, and components (SSCs), and evaluation of the adequacy of defense-in-depth (DID). For advanced reactor designs that contain multiple non-core sources of radioactivity, or for monolithic plant sites that include associated fuel facilities, this approach could potentially simplify licensing applications through the use of a single, uniform, and consistent decision-making framework across all radioactive sources at the site. In addition, a RIPB approach could provide additional insights regarding plant behavior, flexibility regarding licensing decision-making, and potentially allow the use of risk information as part of the plant oversight process. Risk-informing these aspects of advanced reactor licensing would also be consistent with the NRC’s risk policy statement. However, there is diverse regulation and guidance regarding the licensing of non-core sources of radioactivity and generally limited experience using RIPB approaches for the evaluation as part of licensing.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Review and Assessment of Available Data Regarding the Behavior of Sodium Aerosols

Recently, there has been a resurgence of interest in advanced (non-light water) reactor designs, including sodium fast reactors (SFRs). In parallel, multiple efforts are underway to develop risk informed, performance-based licensing pathways for advanced reactors. However, such licensing approaches depend on the accuracy of the associated safety assessments, including mechanistic source term (MST) analyses. Unlike historical source term assessments, which utilizes bounding estimates of radionuclide release, MSTs attempt to realistically estimate the transport and retention of radionuclides for specific transient scenarios. For SFR MST analyses, the behavior of radionuclide aerosols is a key factor, as noted in recent studies. Due to the use of sodium as the primary coolant, many potential transient scenarios involve the release of radionuclide aerosols in conjunction with sodium aerosols. For such scenarios, sodium may be the dominant aerosol species, compared to that of released radionuclides. Therefore, understanding the behavior of sodium aerosols is particularly important to an accurate assessment of aerosol transport. Although many approaches to aerosol modeling exist, it is vital to establish their capabilities in evaluating sodium aerosol behavior for SFR analyses.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA: Structure and Thermophysical Properties of Delta and Epsilon Zry4H x

Small-modular and micro-reactors are part of an emerging nuclear technology aimed to provide power to decentralized grids in civilian, industrial, and defense energy sectors. Resurgence in microreactor research is partly driven by demands for a mobile, safe, and compact energy platform for remote and strategic locations. The defining characteristics of the microreactor include its high-power density on a small footprint, its highly transportable design, and the use of high-assay low-enriched uranium (HALEU) fuel. Solid-state metal hydride neutron moderator components are considered for microreactor designs where the moderator functions to thermalize neutrons into lower energies for the fuel to achieve high fission rates. Zirconium hydride (ZrH x where x denotes the atomic hydrogen-to-metal ratio) is a leading moderator candidate material due to the combination of its low neutron absorption cross section, irradiation resistance, high hydrogen atomic density (upwards of 7.5 x 10 22 H atoms/cm 3 at low temperatures), and high hydrogen desorption (dissociation) temperature around 600 °C (873 K) compared to other metal hydrides. Its high-temperature stability enables operation of fuel temperatures into several hundred degrees Celsius for overall increased efficiency.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Measure the effect of molten halide salt exposure on creep rupture lifetime

Recent resurgence in the research and commercial interests in molten salt reactors (MSRs) as a viable advanced reactor concept to achieve the short- and long-term climate goals has resulted in ongoing efforts to demonstrate their commercial potential. These are relying on a combination of the extensive legacy knowledge from the molten salt reactor experiment (MSRE) and relatively recent data on materials compatibility of structural materials of interest such as 316H in molten salts environments. However, there is a critical lack of data on the mechanical behavior of alloys of interest for MSRS such as 316H, 617 and 709 in molten fluoride (FLiNaK or FLiBe) or chloride (NaCl-MgCl 2 ) salts. Limited legacy data from the molten salt reactor experiment (MSRE) program showed a significant reduction in creep rupture strength of a Ni-base alloy (Ni-15Cr-7Fe wt.%) in the molten fluoride NaF-ZrF4-UF4 (50-46-4 mol.%) salt. With ongoing efforts to commercialize different molten salt reactor concepts, the industry can considerably benefit from quantitative information on the impact of molten halide salts on the engineering properties such as creep and fatigue strength of materials of interest. Creep tests for 316H were conducted with fluoride (FLiNaK) and chloride (NaCl-MgCl 2 ) salts tat 650°C/150 MPa while alloys 709 and 617 were tested with FLiNaK at 700C/158 MPa and 750C.146 MPa respectively. Baseline tests were conducted in air to assess the impact of the molten salts on the creep behavior.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Measure the effect of molten halide salt exposure on creep rupture lifetime

Recent resurgence in the research and commercial interests in molten salt reactors (MSRs) as a viable advanced reactor concept to achieve the short- and long-term climate goals has resulted in ongoing efforts to demonstrate their commercial potential. These are relying on a combination of the extensive legacy knowledge from the molten salt reactor experiment (MSRE) and relatively recent data on materials compatibility of structural materials of interest such as 316H in molten salts environments. However, there is a critical lack of data on the mechanical behavior of alloys of interest for MSRS such as 316H, 617 and 709 in molten fluoride (FLiNaK or FLiBe) or chloride (NaCl-MgCl 2 ) salts. Limited legacy data from the molten salt reactor experiment (MSRE) program showed a significant reduction in creep rupture strength of a Ni-base alloy (Ni-15Cr-7Fe wt.%) in the molten fluoride NaF-ZrF4-UF4 (50-46-4 mol.%) salt. With ongoing efforts to commercialize different molten salt reactor concepts, the industry can considerably benefit from quantitative information on the impact of molten halide salts on the engineering properties such as creep and fatigue strength of materials of interest. Creep tests for 316H were conducted with fluoride (FLiNaK) and chloride (NaCl-MgCl 2 ) salts tat 650°C/150 MPa while alloys 709 and 617 were tested with FLiNaK at 700C/158 MPa and 750C.146 MPa respectively. Baseline tests were conducted in air to assess the impact of the molten salts on the creep behavior.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

2020 Budget Request for the DOE Computational Science Graduate Fellowship (CSGF) Grant

The Department of Energy Computational Science Graduate Fellowship (DOE CSGF) is essential for addressing the increasingly complex national workforce demands stemming from the growth of computational science and engineering challenges. Computational science and engineering (CSE) takes a multidisciplinary approach that utilizes scientific computing to tackle practical problems and provide technical tools across the spectrum of scientific discovery. The DOE CSGF specifically highlights high-performance computing (HPC) as a critical enabling technology in CSE, driving advancements in science and engineering that are vital to both the DOE and the broader economy. Over the past half-century, HPC has been an essential tool for DOE’s success. During this period, important missions, such as nuclear stockpile stewardship, have turned to HPC as an essential technology. Entire science disciplines have been transformed through the augmentation of scientific observation via HPC. At government laboratories, academic institutions, and in industry, DOE CSGF alumni are helping push traditional HPC boundaries while contributing to discoveries in high-energy physics, quantum information systems, fusion-reactor design, machine learning, additive manufacturing, nano materials for next-generation batteries and transistors, and advanced nuclear reactor modeling. In addition, HPC is used to address national health needs that will eventually point to cures both by helping cancer researchers manage and analyze huge troves of data, by simulating biological mechanisms, and by accelerating drug development. A 2023 report from the ASCAC Subcommittee on American Competitiveness and Innovation to the ASCR office, “Can the United States Maintain Its Leadership in High-Performance Computing?” says of the Program, “The CSGF program provides a barometer for disciplines that will be of interest to future DOE computing. Computational biology, machine learning, and quantum computing are among the subjects that began to swell in the ranks of CSGF applicants before the labs were hiring as high a percentage of employees in these categories.” The explosion of scientific and technological data has heightened the demand for advanced high-performance computing (HPC) to transform these data into meaningful scientific insights. As access to vast amounts of data increases, the fields of Machine Learning and Artificial Intelligence are experiencing a resurgence, enhancing the established practices of computational modeling and simulation. In its September 2020 subcommittee report on "AI/ML, Data Intensive Science, and High-Performance Computing," the DOE Advanced Scientific Computing Advisory Committee (ASCAC) specifically called for a fellowship program to train computational and data scientists to address exascale and data-intensive computing challenges. This integration of empirical and theoretical modeling will increasingly guide federal policymakers in making decisions that impact American society and future generations. It demands a workforce of highly skilled and intellectually agile computational scientists capable of navigating the rapid advancements in scientific computing within the DOE National Laboratory research environment. The DOE CSGF program has consistently addressed this critical need.

97 MATHEMATICS AND COMPUTING↗

Progress on the US-Japan Source Term Benchmark Analysis Collaboration

There has been a resurgence in global interest in advanced reactor technology, with a variety of innovative concepts being proposed and developed. Sodium-cooled Fast Reactors (SFRs) have garnered considerable attention given their beneficial characteristics and extensive historical development programs and operating experience. Central to the licensing of non-light water reactor (non-LWR) technology is the characterization of the safety case. At its basics, reactor safety focuses on the prevention and mitigation of the release of radioactive material to the environment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

SPARC - Plans for a New Critical Experiment Facility with a Horizontal Split Table

Several critical experiment facilities, sometimes referred to as zero power reactor facilities, have provided crucial data to aid understanding and validate nuclear-physics models since the beginning of nuclear technology. Indeed, the first man-made reactor, Chicago Pile-1, was essentially this type of reactor. However, there was a downturn in nuclear technology development toward the turn of the millennium, and the need for these specialized research facilities waned. Now there are few of these experimental facilities operational in the world and those that remain have relatively small critical assembly machines. The need for criticality safety benchmark experiments at intermediate neutron energy levels and the modern resurgence of interest in advanced reactors designs, many of which do not have historical precedents in terms of nuclear fuel composition, moderator, and coolant combinations, all combine to create a substantial need for a critical experiment facility with a large horizontal split-table (HST) machine. A HST machine is used to arrange two separate and subcritical parts of a core assembly, bring them together in a precise manner to achieve criticality using remote controls, and separate them to achieve a subcritical configuration again. A new effort was recently performed to develop user needs for a HST, assess candidate locations at the Idaho National Laboratory (INL), and develop a plan for deployment. This project is referred to as the System Physics Advanced Reactor Critical facility (SPARC). A few months after this assessment began, and shortly after as a viable pathway was emerging, a series of important presidential executive orders were issued to revitalize nuclear energy in the United States (U.S.). The relevance of SPARC to these executive orders was immediately apparent. The far-reaching potential of SPARC to these executive orders will reside in its ability to produce data which facilitates licensing of advanced nuclear reactor designs while reducing uncertainties to help increase energy production alongside new criticality safety data to enable more efficient nuclear fuel manufacture, transport, and storage.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

Vertical-Axis Wind Turbine Steady and Unsteady Aerodynamics for Curved Deforming Blades

Vertical-axis wind turbines’ simpler design and low center of gravity make them ideal for floating wind applications. However, efficient design optimization of floating systems requires fast and accurate models. Low-fidelity vertical-axis turbine aerodynamic models, including double multiple streamtube and actuator cylinder theory, were created during the 1980s. Commercial development of vertical-axis turbines all but ceased in the 1990s until around 2010 when interest resurged for floating applications. Despite the age of these models, the original assumptions (2-D, rigid, steady, straight bladed) have not been revisited in full. When the current low-fidelity formulations are applied to modern turbines in the unsteady domain, aerodynamic load errors nearing 50% are found, consistent with prior literature. However, a set of steady and unsteady modifications that remove the majority of error is identified, limiting it near 5%. This paper shows how to reformulate the steady models to allow for unsteady inputs including turbulence, deforming blades, and variable rotational speed. A new unsteady approximation that increases numerical speed by 5–10× is also presented. Combined, these modifications enable full-turbine unsteady simulations with accuracy comparable to higher-fidelity vortex methods, but over 5000× faster.

17 WIND ENERGY↗

Coupled Thermal-Hydraulic Analysis and Species Mass Transport in a Versatile Experimental Salt Irradiation Loop (VESIL) Using CTF

With the resurgence of interest in molten salt reactors, there is a need for new experiments and modeling capabilities to characterize the unique phenomena present in this fluid fuel system. A Versatile Experimental Salt Irradiation Loop (VESIL) is currently under investigation at Idaho National Laboratory to be placed in the Advanced Test Reactor (ATR). One of the key phenomena this proposed experiment plans to elucidate is fission product speciation in the fuel-salt and the subsequent effects this has on the fuel-salt properties, source term generation, and corrosion control. Specifically, noble gases (Xe & Kr) will bubble out to a plenum or off-gas system, and noble metals (Mo, Tc, Te, etc.) will precipitate and deposit in specific zones in the loop. This work extends the mass transfer and species interaction models in CTF (Coolant-Boiling in Rod Arrays—Two Fluids) and applies these models to give a preliminary estimation of fission product behavior in the proposed VESIL design. A noble metal–helium bubble mass transfer model is coupled with the thermal-hydraulic results from CTF to determine the effectiveness of this insoluble fission product (IFP) extraction method for VESIL. Amounts of IFP species extracted to the off-gas system and species distributions in VESIL after a 60-day ATR cycle are reported.

Walker, Samuel A.↗

Solvent-Producing Clostridia Revisited

The review provides an overview of the current status of the solvent-producing clostridia. The origin and development of industrial clostridial species, as well as the history of the industrial Acetone Butanol Ethanol fermentation process, is reexamined, and the recent resurgence of interest in the production of biobutanol is reviewed. Over 300 fully sequenced genomes for solvent-producing and closely related clostridial species are currently available in public databases. These include 270 genomes sourced from the David Jones culture collection. These genomes were allocated arbitrary DJ codes, and a conversion table to identify the species and strains has now been provided. The expanded genomic database facilitated new comparative genomic and phylogenetic analysis. A synopsis of the common features, molecular taxonomy, and phylogeny of solvent-producing clostridia and the application of comparative phylogenomics are evaluated. A survey and analysis of resident prophages in solvent-producing clostridia are discussed, and the discovery, occurrence, and role of novel R-type tailocins are reported. Prophage genomes with R-type tailocin-like features were detected in all 12 species investigated. The widespread occurrence of tailocins in Gram-negative species is well documented; this survey has indicated that they may also be widespread in clostridia.

59 BASIC BIOLOGICAL SCIENCES↗

Coupled Thermal-Hydraulic Analysis and Species Mass Transport in a Versatile Experimental Salt Irradiation Loop (VESIL) using COBRA-TF

With the resurgence of interest in Molten Salt Reactor (MSR) technology among commercial and governmental agencies, there is a need for new experiments and new modelling capabilities to characterize the unique phenomena present in this fluid fuel system. A Versatile Experimental Salt Irradiation Loop (VESIL) is currently under investigation at Idaho National Laboratory. It is designed to be a dynamic fuel-bearing salt loop placed in the Advanced Test Reactor. One of the key phenomena this proposed experiment plans to elucidate is fission product speciation in the fuel-salt and the subsequent effects this can have on fuel-salt properties, fission product transport and interactions, source term generation, corrosion and redox potential control. Specifically, noble gases (Xe & Kr) will bubble out to a plenum or off-gas system and noble metals (Mo, Tc, Te etc.) will precipitate and deposit in specific zones in the loop. This work extends the mass transfer and species interaction models in the general species transport capability of COBRA-TF (CTF) and applies these models to give a preliminary estimation of fission product behavior in the proposed VESIL design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Recovery of Neptunium-236g from Photon and Proton-Irradiated Actinide Targets

Neptunium-236g is a rare radionuclide used as a tracer for neptunium-237 analyses. The availability of 236gNp is limited and the viable production routes are costly, time consuming, and only produce trace quantities of the desired product. For this work, two known production methods were tested to determine product recovery, purity, and viability for use as a tracer. The first method utilizes a photon-irradiated 237Np target to produce 236gNp by the 237Np(?,n)?236Np reaction. The second method utilizes the 238U(d,4n) ?236Np reaction. These production routes were evaluated previously, and the former was considered ineffective without isotope separation and the latter was not well-characterized for the 236mNp/236gNp production ratio. Recent resurgence of electromagnetic isotope separation technology has enabled at least partial recovery of 236gNp from part-per-million abundance feeds produced by the photonuclear reaction. To address the lack of production data for the second method, a deuteron-irradiated depleted uranium target was chemically processed to recover and purify the Np for abundance and ratio analyses. The status and current analytical results for each production method are presented.

07 ISOTOPE AND RADIATION SOURCES↗

The Use of Acoustic Cavitation to Recovery High-Assay Low-Enriched Uranium (HALEU) form TRISO Fuel Particles

Recent interest in advanced nuclear reactor concepts such as small modular reactor (SMR), micro reactor or versatile test reactor (VTR) has increased the demand for high-assay, low-enriched uranium (HALEU). The sustainable management of this resource is integral to successful resurgence of nuclear energy in the United States. The High Temperature Gas-Cooled Reactor (HTGR) concept is a prime candidate for deployment of the new fleet of SMRs, and it will become very important to have a simple and cost-effective technology to recover HALEU from short-burned, defective, or off-specification TRISO fuel particles. We have attempted to harness the high energetics of sonochemistry to penetrate the TRISO fuel particles and recover HALEU via acidic leaching. While the TRISO fuel particle design is inherently “bulletproof,” (due to the silicon carbide (SiC) layer), ultrasonic irradiation may hold the key to convenient access to its valuable contents. Sonochemistry is a field of chemistry based on acoustic cavitation, which is the formation, growth, and collapse of bubbles in liquid media.1,2 The oscillating bubble formation is produced by irradiation of a liquid media with sound waves. Literature reports indicate that collapsing bubbles induced by cavitation produces intense local heating, high pressures, and short lifetimes.1 These localized hot spots reach temperatures of ˜5000 K, pressures approaching 500 atm, and heating and cooling rates exceeding 100 K/s.1 The large temperature and pressure differentials deliver high-energy heating and microscopic explosive shock waves to a liquid media or liquid/solid interface. Cavitation at the surface of a solid in solution induces a deformation in the bubble cavity upon collapse. This deformation reinforces the bubble structure and sends a fast-moving stream of liquid through the cavity at the surface with velocities greater than 100 m/s.3 These energetic impacts have demonstrated an ability to penetrate or simply destroy the SiC shell surrounding the uranium fuel kernel during prolonged exposures to the high-power acoustic waves. Some preliminary results demonstrating HALEU recovery using sonochemistry techniques will be reported.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗