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At least 1,081 records · Page 60

Graphite Specimen Mass Measurements for NESHAP Calculations

The Advanced Graphite Creep (AGC) experiments were designed to irradiate various types of graphite specimens to gain an understanding of how the physical and thermal properties of the graphite change. The graphite specimens are machined from nuclear grade graphite that is very pure relative to other graphite types. However, it still contains parts per million (PPM) levels of metal contaminants that are dispersed throughout it. During irradiation in the Advanced Test Reactor (ATR) these metal contaminants become activated resulting in the graphite specimens being classified as a radioactive material. Following their irradiation, a variety of measurements and experiments are performed on these specimens that require them to be heated above 100°C.Whenever a radioactive material is heated to above 100°C the National Emission Standards for Hazardous Air Pollutants (NESHAP) calculations require the activity content, in Curies, of the isotopes that could potentially be released to the environment. The purpose of this ECAR is to document the method by which the mass loss of each sample is determined and how the isotopic content of this mass loss is determined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Simulated Aging of Metal Tritides by Ball Milling Induced Gas Incorporation

The study of tritium aging effects on materials requires a significant time commitment as a consequence of its 12.3-year half-life, making developmental studies prohibitively difficult and expensive. However, detailed knowledge of long-term aging effects is critical to the development of structural and storage materials for future fusion reactor technologies. As a result, multiple approaches to simulated aging effects have been investigated. Here, we report a method of simulated tritium aging achieved though the incorporation of trapped gases via high-energy ball milling of LaNi 4.25 Al 0.75 alloy storage material. Experimental results verify the presence of trapped gases by a combination of temperature programmed desorption and LECO chemical analysis. Following gas incorporation, we find that many of the degraded hydrogen sorption properties found in aged storage materials are reproduced by the ball milled powders.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Reactor concepts for atomic layer deposition on agitated particles: A review

The number of possible applications for nanoparticles has strongly increased in the last decade. For many applications, nanoparticles with different surface and bulk properties are necessary. A popular surface modification technique is coating the particle surface with a nanometer thick layer. Atomic layer deposition (ALD) is known as a reliable method for depositing ultrathin and conformal coatings. In this article, agitation or fluidization of the particles is necessary for performing ALD on (nano)particles. The principles of gas fluidization of particles will be outlined, and a classification of the gas fluidization behavior of particles based on their size and density will be given. Following different reactor concepts that have been designed to conformally coat (nano)particles with ALD will be described, and a concise overview will be presented of the work that has been performed with each of them ending with a concept reactor for performing spatial ALD on fluidized particles.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

FY17 CAES LDRD Annual Report

Metallic alloys are widely used or planned for use as structural and cladding materials in current and future reactors. Under irradiation, grain boundary (GB) cohesion strength decreases due to interaction with defects and impurities, leading to intergranular fracture and embrittlement of alloys. The objective of this project is to develop a technique for quantifying GB cohesion and its impact on fracture behavior in irradiated alloys, by utilizing transmission electron microscopic (TEM) in situ cantilever testing in concert with multi-scale modeling. The TEM in situ cantilever testing is a novel approach for studying the real-time mechanical response of materials. It will be used in this work for studying intergranular fracture behavior in several irradiated iron-based ferritic alloys and providing key information to link atomistic level events with mesoscale/macroscopic mechanical properties. The Multi-Physics Object-Oriented Simulation Environment (MOOSE)-based cohesive zone model (CZM) and extended finite element method (XFEM) for intergranular fracture of irradiated ferritic alloys will be developed in this work by utilizing atomistic results as inputs and experimental results for validation.

36 MATERIALS SCIENCE↗

Enhancing the quality of radiographic images acquired with point-like gamma-ray sources through correction of the beam divergence and attenuation

Radiographic images acquired with point-like gamma-ray sources exhibit a desirable low penumbra effects specially when positioned far away from the set object-detector. Such an arrangement frequently is not affordable due to the limited flux provided by a distant source. A closer source, however, has two main drawbacks, namely the degradation of the spatial resolution - as actual sources are only approximately punctual - and the non-homogeneity of the beam hitting the detector, which creates a false attenuation map of the object being inspected. This non-homogeneity is caused by the beam divergence itself and by the different thicknesses traversed the beam even if the object were an homogeneous flat plate. In this work, radiographic images of objects with different geometries, such as flat plates and pipes have undergone a correction of beam divergence and attenuation addressing the experimental verification of the capability and soundness of an algorithm formerly developed to generate and process synthetic images. The impact of other parameters, including source-detector gap, attenuation coefficient, ratio defective-to-main hull thickness and counting statistics have been assessed for specifically tailored test-objects aiming at the evaluation of the ability of the proposed method to deal with different boundary conditions. All experiments have been carried out with an X-ray sensitive Imaging Plate and reactor-produced {sup 198}Au and {sup 165}Dy sources. The results have been compared with other technique showing a better capability to correct the attenuation map of inspected objects unveiling their inner structure otherwise concealed by the poor contrast caused by the beam divergence and attenuation, in particular for those regions far apart from the vertical of the source.

07 ISOTOPE AND RADIATION SOURCES↗

Direct Numerical Simulation of Turbulent Flow Inside a Differentially Heated Composite Cavity

In this work, turbulent flow inside a modified differentially heated cavity at high Rayleigh number (Ra ~ 10 9 ) has been studied through fully resolved direct numerical simulation (DNS) using the high-order spectral element method code Nek5000. The flow configuration includes two separate physical phenomena: the natural recirculation itself, and the flow inside a curved channel. Simulations have been carried out using both the Boussinesq approximation and the low-Mach compressible formulation. Significant discrepancies between the two methods inform of the extreme caution that should be exercised when using the Boussinesq approximation in the limits of its applicability. The DNS solutions are analyzed in terms of polynomial-order convergence and Reynolds stress budgets, and the turbulence quantities and velocity profiles are presented as a reference for the validation of turbulence models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Heat Flux Analysis From IR Imaging on Proto-MPEX

The Prototype Material Plasma Exposure eXperiment (Proto-MPEX) is a linear device used to develop the source and heating concept for MPEX, an experiment to study plasma–material interactions for future fusion reactors. This paper highlights recent analysis and results on Proto-MPEX using an infrared (IR) camera, a key tool for imaging heat fluxes measured at the material target. The target heat fluxes are crucial to understanding the source and heating concepts. The recent analysis includes the application of homography and 3-D finite-element methods to provide between-shot measurements of the 2-D target heat flux profile. The IR camera inferred heat flux is shown to correlate with Langmuir probe inferred heat fluxes for a range of parameters in helicon only and helicon + electron cyclotron heated discharges. Results show both desirable central and undesirable edge heat flux for helicon source and electron heating on the target. These heat flux measurements are shown to be important for better understanding of the source and heating physics.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Nuclear fuel alloys or mixtures and method of making thereof

Nuclear fuel alloys or mixtures and methods of making nuclear fuel mixtures are provided. Pseudo-binary actinide-M fuel mixtures form alloys and exhibit: body-centered cubic solid phases at low temperatures; high solidus temperatures; and/or minimal or no reaction or inter-diffusion with steel and other cladding materials. Methods described herein through metallurgical and thermodynamics advancements guide the selection of amounts of fuel mixture components by use of phase diagrams. Weight percentages for components of a metallic additive to an actinide fuel are selected in a solid phase region of an isothermal phase diagram taken at a temperature below an upper temperature limit for the resulting fuel mixture in reactor use. Fuel mixtures include uranium-molybdenum-tungsten, uranium-molybdenum-tantalum, molybdenum-titanium-zirconium, and uranium-molybdenum-titanium systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Bison Verification and Validation Activities for TRISO

Numerical modeling and simulation (M&S) tools play a key role in the research, development, and overall safety assessments of next-generation nuclear energy systems. One such tool, Bison, is a nuclear fuel performance code that is applicable to many fuel forms (e.g., light-water reactor fuel, oxide and metallic fuel for fast reactors, tri-structural isotropic (TRISO) fuel, and plate fuel), and it uses the finite element method to model the thermo- mechanical response of nuclear fuels. One fuel form widely utilized in Generation-IV high-temperature gas-cooled and fluoride- salt-cooled nuclear reactor concepts is TRISO fuel. Recently, Bison’s capabilities were significantly expanded to enable it to model the performance of TRISO particles and compacts. It is important that Bison’s computational results be reliable and predictive, since this code is used to inform high-consequence decisions. The various processes developed to address this issue generally entail two fundamental steps: verification and validation (V&V). Verification ensures that the code functions correctly and is reliable. Code/solution verification, code benchmark, and software quality assurance exercises are examples of verification activities. On the other hand, validation is the process of assessing a code’s capability to accurately model physical problems. Comparisons between code results and experiments quantify the validation level. Application of V&V procedures is crucial to the development of computational tools that are free of coding mistakes and can accurately represent reality. The current study presents an overview of Bison V&V activities relevant to the TRISO fuel concept, which include code/solution verification exercises, CRP-6 Benchmark—a Coordinated Research Program through the International Atomic Energy Agency (IAEA)—exercises, and validation exercises with the Advanced Gas Reactor (AGR)- 1/2/3/4 experiment series.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Development of Seismic Isolation Systems Using Periodic Materials

Advanced fast nuclear power plants and small modular fast reactors are composed of thin-walled structures such as pipes; as a result, they do not have sufficient inherent strength to resist seismic loads. Seismic isolation, therefore, is an effective solution for mitigating earthquake hazards for these types of structures. Base isolation, on which numerous studies have been conducted, is a well-defined structure protection system against earthquakes. In conventional isolators, such as high-damping rubber bearings, lead-rubber bearings, and friction pendulum bearings, large relative displacements occur between upper structures and foundations. Only isolation in a horizontal direction is provided; these features are not desirable for the piping systems. The concept of periodic materials, based on the theory of solid-state physics, can be applied to earthquake engineering. The periodic material is a material that possesses distinct characteristics that prevent waves with certain frequencies from being transmitted through it; therefore, this material can be used in structural foundations to block unwanted seismic waves with certain frequencies. The frequency band of periodic material that can filter out waves is called the band gap, and the structural foundation made of periodic material is referred to as the periodic foundation. The design of a nuclear power plant, therefore, can be unified around the desirable feature of a periodic foundation, while the continuous maintenance of the structure is not needed. In this research project, three different types of periodic foundations were studied: one-dimensional, two-dimensional, and three-dimensional. The basic theories of periodic foundations are introduced first to find the band gaps; then the finite element methods are used, to perform parametric analysis, and obtain attenuation zones; finally, experimental programs are conducted, and the test data are analyzed to verify the theory. This procedure shows that the periodic foundation is a promising and effective way to mitigate structural damage caused by earthquake excitation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Deployment of Traditional and Hybrid Machine Learning for Critical Heat Flux Prediction in the CTF Thermal-Hydraulics Code

Critical heat flux (CHF) marks the transition from nucleate to film boiling, where heat transfer to the working fluid can rapidly deteriorate. Accurate CHF prediction is essential for efficiency, safety, and preventing equipment damage, particularly in nuclear reactors. Although widely used, empirical correlations frequently exhibit discrepancies when compared to experimental data, limiting their reliability in diverse operational conditions. Traditional machine learning (ML) approaches have demonstrated potential for CHF prediction but often suffer from limited interpretability, data scarcity, and insufficient knowledge of physical principles. Hybrid model approaches, which combine data-driven ML with base models, mitigate these concerns by incorporating prior knowledge of the domain. This study integrates an externally trained purely data-driven ML model and two hybrid models (using the Biasi and Bowring CHF correlations) within the CTF subchannel code via a custom Fortran framework. Performance was evaluated using two validation cases: a subset of the Nuclear Regulatory Commission (NRC) CHF database and the Bennett dryout experiments. In both cases, the hybrid models demonstrated significantly lower error metrics compared to conventional empirical correlations, with the best models often reducing relative error by about 5 percentage points. The pure ML model achieved comparable accuracy, outperforming the hybrid Biasi model in the NRC test case (3.3% versus 5.5% relative error) but exhibiting slightly higher error against the hybrid Bowring model in the Bennett test case (7.7% versus 6.1%). Trend analysis of error parity indicated that ML-based models reduced the tendency for CHF overprediction, improving overall accuracy. These results demonstrate that ML-based CHF models can be effectively integrated into subchannel codes and could potentially increase performance compared to conventional methods.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

An innovative and advanced coupled neutron transport and thermal hydraulic method (tool) for the design, analysis and optimization of VHTR/NGNP prismatic reactors

This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implement a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Production of chemicals and fuels from biomass

Methods, reactor systems, and catalysts are provided for converting in a continuous process biomass to fuels and chemicals, including methods of converting the water insoluble components of biomass, such as hemicellulose, cellulose and lignin, to volatile C.sub.2+O.sub.1-2 oxygenates, such as alcohols, ketones, cyclic ethers, esters, carboxylic acids, aldehydes, and mixtures thereof. In certain applications, the volatile C.sub.2+O.sub.1-2 oxygenates can be collected and used as a final chemical product, or used in downstream processes to produce liquid fuels, chemicals and other products.

09 BIOMASS FUELS↗

Silicon carbide reinforced zirconium based cladding

A method for making an improved nuclear fuel cladding tube includes reinforcing a Zr alloy tube by first winding or braiding ceramic yarn directly around the tube to form a ceramic covering, then physically bonding the ceramic covering to the tube by applying a first coating selected from the group consisting of Nb, Nb alloy, Nb oxide, Cr, Cr oxide, Cr alloy, or combinations thereof, by one of a thermal deposition process or a physical deposition process to provide structural support member for the Zr tube, and optionally applying a second coating and optionally applying a third coating by one of a thermal deposition process or a physical deposition process. If the tube softens at 800° C.-1000° C., the structural support tube will reinforce the Zr alloy tube against ballooning and bursting, thereby preventing the release of fission products to the reactor coolant.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

A Comparison of Modeling Strategies for Additively Manufactured HFIR Control Elements

The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory is a flux-trap type research reactor that is used for a range of research applications, including neutron scattering experiments, materials irradiation tests, and isotope production. The HFIR core comprises several concentric cylindrical regions; starting from the core centerline and moving radially outwards, these are the flux trap target (FTT) region, the inner and outer fuel elements (IFE and OFE), the inner and outer control elements (ICE and OCE), and the reflector region. The control elements (CEs) are axially zoned into regions of differing absorption strength: the white regions contain only structural aluminum and do not significantly contribute to reactivity control; the gray regions contain tantalum as the primary absorber with a concentration of 38% by volume, and the black regions contain europia (Eu{sub 2}O{sub 3}) as the primary absorber with a concentration of 31% by volume. During operation, the CEs are withdrawn symmetrically in opposite directions from the core to remove the neutron absorbers from the core and maintain criticality. The current approach to fabrication of the HFIR CEs is tedious, outdated, and expensive. In an effort to improve the operational economics of HFIR and to demonstrate the feasibility of advanced manufacturing techniques for production of nuclear reactor components, novel methods for production of neutron-poison bearing aluminum structures have been initiated at ORNL. For this application, the ultrasonic additive manufacturing (UAM) process involves the use of 3D printing technology to construct small cylindrical compacts (0.559 cm in diameter and 0.171 cm in height) of aluminum mixed with neutron-absorbing material. The compacts contain 60% poison (Ta or Eu{sub 2}O{sub 3}) by volume. These compacts are then embedded into an aluminum matrix before formation into the cylindrical shape appropriate for use in HFIR. The CEs produced in this process therefore contain discrete regions of absorbing material, in contrast to the traditional design with homogeneously distributed absorbers. It is therefore of interest to investigate how the HFIR core behavior with discrete CEs differs from the case with homogeneous CEs with respect to some key reactor physics parameters. The primary objective of this study is to compare modeling strategies for the discrete CE design and examine differences in the core physics between the discrete and homogeneous CE designs as predicted by each model. Lessons learned from this investigation may then be used to inform the design process in future endeavors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Major Achievements of the Springfields Decommissioning Program - 16512

Since the 1940's, Springfields has manufactured fuel products for the UK's nuclear power stations and for international customers. The early operations were in support of the atomic weapons program, closely followed by fuel manufacture for electricity generation via the Magnox power stations commencing in the late 1950's. In 2005, responsibility for the assets and liabilities of Springfields was transferred to the Nuclear Decommissioning Authority (NDA). A new company, Springfields Fuels Limited, was created to run the site, which was managed and operated by Westinghouse Electric UK Ltd on the NDA's behalf. On April 1, 2010, Westinghouse entered into an agreement with the NDA for a long-term lease of the Springfields site, which transferred responsibility for the commercial fuel manufacturing business and Springfields Fuels Limited to Westinghouse. Today, the main Springfields activities include: - Oxide fuels for Advanced Gas-cooled and Light Water Reactors, as well as intermediate fuel products, such as powders, granules and pellets, - Manufacture of Uranium Hexafluoride (now in Care and Maintenance), - Processing and recovery of Uranium residues, - Decommissioning and demolition of redundant plants and buildings. Some 120 plant buildings and support facilities have been successfully decommissioned and demolished in the period 1996 to date. These range from 1940's buildings, subsequently converted for nuclear fuel manufacture, together with some of the formerly sealed chemical agent manufacture and storage facilities. Decommissioning of the now redundant Magnox manufacturing facilities is ongoing at Springfields without compromising the other site operations. Westinghouse is managing all activities related to the whole lifecycle, planning, design, safety case, radiation protection, Post Operational Clean Out (POCO), service diversions, characterization, decommissioning all phases through to ground remediation, waste disposal and agreed end states. Decommissioning of process plants starts with the removal of the majority of radiological activity or hazard contained within the plant. This is known as Post Operational Clean Out. Once this is complete, the plant may be dismantled. Work is scoped and defined within packages (at plant or cell level) to aid control and commercial monitoring as the decommissioning is progressed. This package approach also enables clean or lightly contaminated items to be removed separately to aid potential re-cycling and minimize decontamination and waste disposal costs. Westinghouse is further optimising work package delivery by planning parallel POCO and Decommissioning work phases within plant at a cell level to systematically remove redundant plant and reduce hazard earlier, whilst optimising use of both internal and external resources to reduce total project cost. Waste and Residues from all phases are disposed of via authorized disposal routes and in accordance with our environmental authorizations. On completion of decommissioning and removal of plant and equipment the building is re-assessed radiologically, categorized and appropriate remediation work progressed to render it safe for conventional demolition. The major elements of the success achieved to date can be attributed to the following elements: - Overall programme management with extensive experience of planning, cost, and risk management. Integration of the programme within a major commercial site without compromising safety or operations. - Complete site EHS management working within a complex regulatory framework including the Nuclear Site Licenses, Environmental Authorisations and Security requirements. - Avoiding safety complacency by use of rigorous and systematic process to emphasise safety over competing goals. Including structured use of Human Performance and Corrective Action and Learning processes. - Unique safety experiences attained through complex hazard reduction across a broad range of uranium manufacturing and chemical hazard/agent facilities. - Project leadership and excellence utilizing a core site team to deploy repeatable processes and safe practices to deliver complex scope within budget and schedule. - Optimised use of internal and external resources, with extensive use of Operating Experience to drive continuous improvement and problem resolution. - Management of Contractors from initial selection, pre-qualification, tender assessment and Performance using varied contract forms across delivery phases. - Use of Westinghouse expertise for safety case, waste management, uranium recovery, health physics and design. - Implementation of a work package to ensure gated discreet manageable packages. Underpinned by package specific method statements, with appropriate hold points to maintain management and control of the works. - Continuous local safety focus via safety planning, induction, plant specific training, proactive risk assessment, and a defined approach to decoupling operating plant safety cases enabling focussed and usable documentation. This paper will provide an overview of the latest decommissioning activities achieved on the Springfields site and will focus on best practices that allowed Westinghouse to maintain the program on time and on budget. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Synthesis of Hexagonal Boron Nitride Mono layer: Control of Nucleation and Crystal Morphology

Mono layer hexagonal boron nitride (hBN) attracts significant attention due to the potential to be used as a complementary two-dimensional dielectric in fabrication of functional 2D heterostructures. Here we investigate the growth stages of the hBN single crystals and show that hBN crystals change their shape from triangular to truncated triangular and further to hexagonal depending on copper substrate distance from the precursor. We suggest that the observed hBN crystal shape variation is affected by the ratio of boron to nitrogen active species concentrations on the copper surface inside the CVD reactor. Strong temperature dependence reveals the activation energies for the hBN nucleation process of similar to 5 eV and crystal growth of similar to 3.5 eV. We also show that the resulting h-BN film morphology is strongly affected by the heating method of borazane precursor and the buffer gas. Elucidation of these details facilitated synthesis of high quality large area monolayer hexagonal boron nitride by atmospheric pressure chemical vapor deposition on copper using borazane as a precursor.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Measurement of the Kinematic Viscosity of Melted Mixtures of Sodium, Lithium, and Beryllium Fluorides and the Effect of the Eutectic Additive Cerium Trifluoride on the Viscosity

Experimental data on the kinematic viscosity and liquidus temperature of certain fused mixtures of sodium, lithium, and beryllium fluorides were obtained, by the method of damping of torsional oscillations of a cylinder with the experimental melt, in validation of the concept of molten-salt incinerator reactors for incinerating transuranium elements from spent nuclear fuel from light-water reactors. The measurements were conducted from the liquidus temperature to 800°C. The effect of the addition of cerium trifluoride on the viscosity of the eutectic melt 58NaF–15LiF–27BeF{sub 2} was studied. It was shown that the additive appreciably reduces the viscosity at low temperature and also lowers the liquidus temperature. The parameters characterizing the temperature dependence of viscosity were determined for the experimental melts.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗