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Improvements to Nuclear Data and Its Uncertainties by Theoretical Modeling

Description: This project addresses three important gaps in existing evaluated nuclear data libraries that represent a significant hindrance against highly advanced modeling and simulation capabilities for the Advanced Fuel Cycle Initiative (AFCI). This project will: Develop advanced theoretical tools to compute prompt fission neutrons and gamma-ray characteristics well beyond average spectra and multiplicity, and produce new evaluated files of U and Pu isotopes, along with some minor actinides; Perform state-of-the-art fission cross-section modeling and calculations using global and microscopic model input parameters, leading to truly predictive fission cross-sections capabilities. Consistent calculations for a suite of Pu isotopes will be performed; Implement innovative data assimilation tools, which will reflect the nuclear data evaluation process much more accurately, and lead to a new generation of uncertainty quantification files. New covariance matrices will be obtained for Pu isotopes and compared to existing ones. The deployment of a fleet of safe and efficient advanced reactors that minimize radiotoxic waste and are proliferation-resistant is a clear and ambitious goal of AFCI. While in the past the design, construction and operation of a reactor were supported through empirical trials, this new phase in nuclear energy production is expected to rely heavily on advanced modeling and simulation capabilities. To be truly successful, a program for advanced simulations of innovative reactors will have to develop advanced multi-physics capabilities, to be run on massively parallel super- computers, and to incorporate adequate and precise underlying physics. And all these areas have to be developed simultaneously to achieve those ambitious goals. Of particular interest are reliable fission cross-section uncertainty estimates (including important correlations) and evaluations of prompt fission neutrons and gamma-ray spectra and uncertainties.
Date: February 18, 2013
Creator: Danon, Yaron; Nazarewicz, Witold & Talou, Patrick
Partner: UNT Libraries Government Documents Department

Modeling Fission Product Sorption in Graphite Structures

Description: The goal of this project is to determine changes in adsorption and desorption of fission products to/from nuclear-grade graphite in response to a changing chemical environment. First, the project team will employ principle calculations and thermodynamic analysis to predict stability of fission products on graphite in the presence of structural defects commonly observed in very high- temperature reactor (VHTR) graphites. Desorption rates will be determined as a function of partial pressure of oxygen and iodine, relative humidity, and temperature. They will then carry out experimental characterization to determine the statistical distribution of structural features. This structural information will yield distributions of binding sites to be used as an input for a sorption model. Sorption isotherms calculated under this project will contribute to understanding of the physical bases of the source terms that are used in higher-level codes that model fission product transport and retention in graphite. The project will include the following tasks: Perform structural characterization of the VHTR graphite to determine crystallographic phases, defect structures and their distribution, volume fraction of coke, and amount of sp2 versus sp3 bonding. This information will be used as guidance for ab initio modeling and as input for sorptivity models; Perform ab initio calculations of binding energies to determine stability of fission products on the different sorption sites present in nuclear graphite microstructures. The project will use density functional theory (DFT) methods to calculate binding energies in vacuum and in oxidizing environments. The team will also calculate stability of iodine complexes with fission products on graphite sorption sites; Model graphite sorption isotherms to quantify concentration of fission products in graphite. The binding energies will be combined with a Langmuir isotherm statistical model to predict the sorbed concentration of fission products on each type of graphite site. The model will include multiple simultaneous adsorbing ...
Date: April 8, 2013
Creator: Szlufarska, Izabela; Morgan, Dane & Allen, Todd
Partner: UNT Libraries Government Documents Department

A New 2D-Transport, 1D-Diffusion Approximation of the Boltzmann Transport equation

Description: The work performed in this project consisted of the derivation, implementation, and testing of a new, computationally advantageous approximation to the 3D Boltz- mann transport equation. The solution of the Boltzmann equation is the neutron flux in nuclear reactor cores and shields, but solving this equation is difficult and costly. The new “2D/1D” approximation takes advantage of a special geometric feature of typical 3D reactors to approximate the neutron transport physics in a specific (ax- ial) direction, but not in the other two (radial) directions. The resulting equation is much less expensive to solve computationally, and its solutions are expected to be sufficiently accurate for many practical problems. In this project we formulated the new equation, discretized it using standard methods, developed a stable itera- tion scheme for solving the equation, implemented the new numerical scheme in the MPACT code, and tested the method on several realistic problems. All the hoped- for features of this new approximation were seen. For large, difficult problems, the resulting 2D/1D solution is highly accurate, and is calculated about 100 times faster than a 3D discrete ordinates simulation.
Date: June 17, 2013
Creator: Larsen, Edward
Partner: UNT Libraries Government Documents Department

Preparation of High Purity, High Molecular-Weight Chitin from Ionic Liquids for Use as an Adsorbate for the Extraction of Uranium from Seawater (Workscope MS-FC: Fuel Cycle R&D)

Description: Ensuring a domestic supply of uranium is a key issue facing the wider implementation of nuclear power. Uranium is mostly mined in Kazakhstan, Australia, and Canada, and there are few high-grade uranium reserves left worldwide. Therefore, one of the most appealing potential sources of uranium is the vast quantity dissolved in the oceans (estimated to be 4.4 billion tons worldwide). There have been research efforts centered on finding a means to extract uranium from seawater for decades, but so far none have resulted in an economically viable product, due in part to the fact that the materials that have been successfully demonstrated to date are too costly (in terms of money and energy) to produce on the necessary scale. Ionic Liquids (salts which melt below 100{degrees}C) can completely dissolve raw crustacean shells, leading to recovery of a high purity, high molecular weight chitin powder and to fibers and films which can be spun directly from the extract solution suggesting that continuous processing might be feasible. The work proposed here will utilize the unprecedented control this makes possible over the chitin fiber a) to prepare electrospun nanofibers of very high surface area and in specific architectures, b) to modify the fiber surfaces chemically with selective extractant capacity, and c) to demonstrate their utility in the direct extraction and recovery of uranium from seawater. This approach will 1) provide direct extraction of chitin from shellfish waste thus saving energy over the current industrial process for obtaining chitin; 2) allow continuous processing of nanofibers for very high surface area fibers in an economical operation; 3) provide a unique high molecular weight chitin not available from the current industrial process leading to stronger, more durable fibers; and 4) allow easy chemical modification of the large surface areas of the fibers for appending uranyl selective ...
Date: December 21, 2013
Creator: Rogers, Robin
Partner: UNT Libraries Government Documents Department

Fundamental Understanding of Ambient and High-Temperature Plasticity Phenomena in Structural Materials in Advanced Reactors

Description: The goal of this research project is to develop the methods and tools necessary to link unit processes analyzed using atomistic simulations involving interaction of vacancies and interstitials with dislocations, as well as dislocation mediation at sessile junctions and interfaces as affected by radiation, with cooperative influence on higher-length scale behavior of polycrystals. These tools and methods are necessary to design and enhance radiation-induced damage-tolerant alloys. The project will achieve this goal by applying atomistic simulations to characterize unit processes of: 1. Dislocation nucleation, absorption, and desorption at interfaces 2. Vacancy production, radiation-induced segregation of substitutional Cr at defect clusters (point defect sinks) in BCC Fe-Cr ferritic/martensitic steels 3. Investigation of interaction of interstitials and vacancies with impurities (V, Nb, Ta, Mo, W, Al, Si, P, S) 4. Time evolution of swelling (cluster growth) phenomena of irradiated materials 5. Energetics and kinetics of dislocation bypass of defects formed by interstitial clustering and formation of prismatic loops, informing statistical models of continuum character with regard to processes of dislocation glide, vacancy agglomeration and swelling, climb and cross slip This project will consider the Fe, Fe-C, and Fe-Cr ferritic/martensitic material system, accounting for magnetism by choosing appropriate interatomic potentials and validating with first principles calculations. For these alloys, the rate of swelling and creep enhancement is considerably lower than that of face-centered cubic (FCC) alloys and of austenitic Fe-Cr-Mo alloys. The team will confirm mechanisms, validate simulations at various time and length scales, and improve the veracity of computational models. The proposed research?s feasibility is supported by recent modeling of radiation effects in metals and alloys, interfacial dislocation transfer reactions in nano-twinned copper, and dislocation reactions at general boundaries, along with extensive modeling cooperative effects of dislocation interactions and migration in crystals and polycrystals using continuum models.
Date: November 17, 2013
Creator: Deo, Chaitanya; Zhu, Ting & McDowell, David
Partner: UNT Libraries Government Documents Department

Investigation of a Novel NDE Method for Monitoring Thermomechanical Damage and Microstructure Evolution in Ferritic-Martensitic Steels for Generation IV Nuclear Energy Systems

Description: The main goal of the proposed project is the development of validated nondestructive evaluation (NDE) techniques for in situ monitoring of ferritic-martensitic steels like Grade 91 9Cr-1Mo, which are candidate materials for Generation IV nuclear energy structural components operating at temperatures up to ~650{degree}C and for steam-generator tubing for sodium-cooled fast reactors. Full assessment of thermomechanical damage requires a clear separation between thermally activated microstructural evolution and creep damage caused by simultaneous mechanical stress. Creep damage can be classified as "negligible" creep without significant plastic strain and "ordinary" creep of the primary, secondary, and tertiary kind that is accompanied by significant plastic deformation and/or cavity nucleation and growth. Under negligible creep conditions of interest in this project, minimal or no plastic strain occurs, and the accumulation of creep damage does not significantly reduce the fatigue life of a structural component so that low-temperature design rules, such as the ASME Section III, Subsection NB, can be applied with confidence. The proposed research project will utilize a multifaceted approach in which the feasibility of electrical conductivity and thermo-electric monitoring methods is researched and coupled with detailed post-thermal/creep exposure characterization of microstructural changes and damage processes using state-of-the-art electron microscopy techniques, with the aim of establishing the most effective nondestructive materials evaluation technique for particular degradation modes in high-temperature alloys that are candidates for use in the Next Generation Nuclear Plant (NGNP) as well as providing the necessary mechanism-based underpinnings for relating the two. Only techniques suitable for practical application in situ will be considered. As the project evolves and results accumulate, we will also study the use of this technique for monitoring other GEN IV materials. Through the results obtained from this integrated materials behavior and NDE study, new insight will be gained into the best nondestructive creep and microstructure monitoring methods ...
Date: September 30, 2013
Creator: Nagy, Peter
Partner: UNT Libraries Government Documents Department

Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

Description: 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.
Date: November 29, 2013
Creator: Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi & Zhang, Dingkang
Partner: UNT Libraries Government Documents Department

Implementation of On-the-Fly Doppler Broadening in MCNP5 for Multiphysics Simulation of Nuclear Reactors

Description: A new method to obtain Doppler broadened cross sections has been implemented into MCNP, removing the need to generate cross sections for isotopes at problem temperatures. Previous work had established the scientific feasibility of obtaining Doppler-broadened cross sections "on-the-fly" (OTF) during the random walk of the neutron. Thus, when a neutron of energy E enters a material region that is at some temperature T, the cross sections for that material at the exact temperature T are immediately obtained by interpolation using a high order functional expansion for the temperature dependence of the Doppler-broadened cross section for that isotope at the neutron energy E. A standalone Fortran code has been developed that generates the OTF library for any isotope that can be processed by NJOY. The OTF cross sections agree with the NJOY-based cross sections for all neutron energies and all temperatures in the range specified by the user, e.g., 250K - 3200K. The OTF methodology has been successfully implemented into the MCNP Monte Carlo code and has been tested on several test problems by comparing MCNP with conventional ACE cross sections versus MCNP with OTF cross sections. The test problems include the Doppler defect reactivity benchmark suite and two full-core VHTR configurations, including one with multiphysics coupling using RELAP5-3D/ATHENA for the thermal-hydraulic analysis. The comparison has been excellent, verifying that the OTF libraries can be used in place of the conventional ACE libraries generated at problem temperatures. In addition, it has been found that using OTF cross sections greatly reduces the complexity of the input for MCNP, especially for full-core temperature feedback calculations with many temperature regions. This results in an order of magnitude decrease in the number of input lines for full-core configurations, thus simplifying input preparation and reducing the potential for input errors. Finally, for full-core problems with ...
Date: November 16, 2012
Creator: Martin, William
Partner: UNT Libraries Government Documents Department

Modeling Solute Thermokinetics in LiCI-KCI Molten Salt for Nuclear Waste Separation

Description: Recovery of actinides is an integral part of a closed nuclear fuel cycle. Pyrometallurgical nuclear fuel recycling processes have been developed in the past for recovering actinides from spent metallic and nitride fuels. The process is essentially to dissolve the spent fuel in a molten salt and then extract just the actinides for reuse in a reactor. Extraction is typically done through electrorefining, which involves electrochemical reduction of the dissolved actinides and plating onto a cathode. Knowledge of a number of basic thermokinetic properties of salts and salt-fuel mixtures is necessary for optimizing present and developing new approaches for pyrometallurgical waste processing. The properties of salt-fuel mixtures are presently being studied, but there are so many solutes and varying concentrations that direct experimental investigation is prohibitively time consuming and expensive (particularly for radioactive elements like Pu). Therefore, there is a need to reduce the number of required experiments through modeling of salt and salt-fuel mixture properties. This project will develop first-principles-based molecular modeling and simulation approaches to predict fundamental thermokinetic properties of dissolved actinides and fission products in molten salts. The focus of the proposed work is on property changes with higher concentrations (up to 5 mol%) of dissolved fuel components, where there is still very limited experimental data. The properties predicted with the modeling will be density, which is used to assess the amount of dissolved material in the salt; diffusion coefficients, which can control rates of material transport during separation; and solute activity, which determines total solubility and reduction potentials used during electrorefining. The work will focus on La, Sr, and U, which are chosen to include the important distinct categories of lanthanides, alkali earths, and actinides, respectively. Studies will be performed using LiCl-KCl salt at the eutectic composition (58 mol% LiCl, 42 mol% KCl), which is used ...
Date: October 1, 2013
Creator: Morgan, Dane & Eapen, Jacob
Partner: UNT Libraries Government Documents Department

Modeling Stress Strain Relationships and Predicting Failure Probabilities For Graphite Core Components

Description: This project will implement inelastic constitutive models that will yield the requisite stress-strain information necessary for graphite component design. Accurate knowledge of stress states (both elastic and inelastic) is required to assess how close a nuclear core component is to failure. Strain states are needed to assess deformations in order to ascertain serviceability issues relating to failure, e.g., whether too much shrinkage has taken place for the core to function properly. Failure probabilities, as opposed to safety factors, are required in order to capture the bariability in failure strength in tensile regimes. The current stress state is used to predict the probability of failure. Stochastic failure models will be developed that can accommodate possible material anisotropy. This work will also model material damage (i.e., degradation of mechanical properties) due to radiation exposure. The team will design tools for components fabricated from nuclear graphite. These tools must readily interact with finite element software--in particular, COMSOL, the software algorithm currently being utilized by the Idaho National Laboratory. For the eleastic response of graphite, the team will adopt anisotropic stress-strain relationships available in COMSO. Data from the literature will be utilized to characterize the appropriate elastic material constants.
Date: September 9, 2013
Creator: Duffy, Stephen
Partner: UNT Libraries Government Documents Department

Chalcogenide Glass Radiation Sensor; Materials Development, Design and Device Testing

Description: For many decades, various radiation detecting material have been extensively researched, to find a better material or mechanism for radiation sensing. Recently, there is a growing need for a smaller and effective material or device that can perform similar functions of bulkier Geiger counters and other measurement options, which fail the requirement for easy, cheap and accurate radiation dose measurement. Here arises the use of thin film chalcogenide glass, which has unique properties of high thermal stability along with high sensitivity towards short wavelength radiation. The unique properties of chalcogenide glasses are attributed to the lone pair p-shell electrons, which provide some distinctive optical properties when compared to crystalline material. These qualities are derived from the energy band diagram and the presence of localized states in the band gap. Chalcogenide glasses have band tail states and localized states, along with the two band states. These extra states are primarily due to the lone pair electrons as well as the amorphous structure of the glasses. The localized states between the conductance band (CB) and valence band (VB) are primarily due to the presence of the lone pair electrons, while the band tail states are attributed to the Van der Waal’s forces between layers of atoms [1]. Localized states are trap locations within the band gap where electrons from the valence band can hop into, in their path towards the conduction band. Tail states on the other hand are locations near the band gap edges and are known as Urbach tail states (Eu). These states are occupied with many electrons that can participate in the various transformations due to interaction with photons. According to Y. Utsugi et. al.[2], the electron-phonon interactions are responsible for the generation of the Urbach tails. These states are responsible for setting the absorption edge for these glasses ...
Date: April 30, 2013
Creator: Mitkova, Maria; Butt, Darryl; Kozicki, Michael & Barnaby, Hugo
Partner: UNT Libraries Government Documents Department

Assessment of Embrittlement of VHTR Structural Alloys in Impure Helium Environments

Description: The helium coolant in high-temperature reactors inevitably contains low levels of impurities during steady-state operation, primarily consisting of small amounts of H{sub 2}, H{sub 2}O, CH{sub 4}, CO, CO{sub 2}, and N{sub 2} from a variety of sources in the reactor circuit. These impurities are problematic because they can cause significant long-term corrosion in the structural alloys used in the heat exchangers at elevated temperatures. Currently, the primary candidate materials for intermediate heat exchangers are Alloy 617, Haynes 230, Alloy 800H, and Hastelloy X. This project will evaluate the role of impurities in helium coolant on the stress-assisted grain boundary oxidation and creep crack growth in candidate alloys at elevated temperatures. The project team will: • Evaluate stress-assisted grain boundary oxidation and creep crack initiation and crack growth in the temperature range of 500-850°C in a prototypical helium environment. • Evaluate the effects of oxygen partial pressure on stress-assisted grain boundary oxidation and creep crack growth in impure helium at 500°C, 700°C, and 850°C respectively. • Characterize the microstructure of candidate alloys after long-term exposure to an impure helium environment in order to understand the correlation between stress-assisted grain boundary oxidation, creep crack growth, material composition, and impurities in the helium coolant. • Evaluate grain boundary engineering as a method to mitigate stress-assisted grain boundary oxidation and creep crack growth of candidate alloys in impure helium. The maximum primary helium coolant temperature in the high-temperature reactor is expected to be 850-1,000°C.Corrosion may involve oxidation, carburization, or decarburization mechanisms depending on the temperature, oxygen partial pressure, carbon activity, and alloy composition. These corrosion reactions can substantially affect long-term mechanical properties such as crack- growth rate and fracture toughness, creep rupture, and fatigue. Although there are some studies on the effects of impurities in helium coolant on creep rupture and fatigue strength, ...
Date: May 31, 2013
Creator: Crone, Wendy; Cao, Guoping & Sridhara, Kumar
Partner: UNT Libraries Government Documents Department

Advanced Models of LWR Pressure Vessel Embrittlement for Low Flux-HighFluence Conditions

Description: Neutron embrittlement of reactor pressure vessels (RPVs) is an unresolved issue for light water reactor life extension, especially since transition temperature shifts (TTS) must be predicted for high 80-year fluence levels up to approximately 1,020 n/cm{sup 2}, far beyond the current surveillance database. Unfortunately, TTS may accelerate at high fluence, and may be further amplified by the formation of late blooming phases that result in severe embrittlement even in low-copper (Cu) steels. Embrittlement by this mechanism is a potentially significant degradation phenomenon that is not predicted by current regulatory models. This project will focus on accurately predicting transition temperature shifts at high fluence using advanced physically based, empirically validated and calibrated models. A major challenge is to develop models that can adjust test reactor data to account for flux effects. Since transition temperature shifts depend on synergistic combinations of many variables, flux-effects cannot be treated in isolation. The best current models systematically and significantly under-predict transition temperature at high fluence, although predominantly for irradiations at much higher flux than actual RPV service. This project will integrate surveillance, test reactor and mechanism data with advanced models to address a number of outstanding RPV embrittlement issues. The effort will include developing new databases and preliminary models of flux effects for irradiation conditions ranging from very low (e.g., boiling water reactor) to high (e.g., accelerated test reactor). The team will also develop a database and physical models to help predict the conditions for the formation of Mn-Ni-Si late blooming phases and to guide future efforts to fully resolve this issue. Researchers will carry out other tasks on a best-effort basis, including prediction of transition temperature shift attenuation through the vessel wall, remediation of embrittlement by annealing, and fracture toughness master curve issues.
Date: June 17, 2013
Creator: Odette, G. Robert & Yamamoto, Takuya
Partner: UNT Libraries Government Documents Department

Development of Alternative Technetium Waste Forms

Description: The UREX+1 process is under consideration for the separation of transuranic elements from spent nuclear fuel. The first steps of this process extract the fission product technicium-99 ({sup 99}Tc) into an organic phase containing tributylphosphate together with uranium. Treatment of this stream requires the separation of Tc from U and placement into a suitable waste storage form. A potential candidate waste form involves immobilizing the Tc as an alloy with either excess metallic zirconium or stainless steel. Although Tc-Zr alloys seem to be promising waste forms, alternative materials must be investigated. Innovative studies related to the synthesis and behavior of a different class of Tc materials will increase the scientific knowledge related to development of Tc waste forms. These studies will also provide a better understanding of the behavior of {sup 99}Tc in repository conditions. A literature survey has selected promising alternative waste forms for further study: technetium metallic alloys, nitrides, oxides, sulfides, and pertechnetate salts. The goals of this project are to 1) synthesize and structurally characterize relevant technetium materials that may be considered as waste forms, 2) investigate material behavior in solution under different conditions of temperature, electrochemical potential, and radiation, and 3) predict the long-term behavior of these materials.
Date: September 13, 2013
Creator: Czerwinski, Kenneth
Partner: UNT Libraries Government Documents Department

Study of Interfacial Interactions Using Thing Film Surface Modification: Radiation and Oxidation Effects in Materials

Description: Interfaces play a key role in dictating the long-term stability of materials under the influence of radiation and high temperatures. For example, grain boundaries affect corrosion by way of providing kinetically favorable paths for elemental diffusion, but they can also act as sinks for defects and helium generated during irradiation. Likewise, the retention of high-temperature strength in nanostructured, oxide-dispersion strengthened steels depends strongly on the stoichiometric and physical stability of the (Y, Ti)-oxide particles/matrix interface under radiation and high temperatures. An understanding of these interfacial effects at a fundamental level is important for the development of materials for extreme environments of nuclear reactors. The goal of this project is to develop an understanding stability of interfaces by depositing thin films of materials on substrates followed by ion irradiation of the film-substrate system at elevated temperatures followed by post-irradiation oxidation treatments. Specifically, the research will be performed by depositing thin films of yttrium and titanium (~500 nm) on Fe-12%Cr binary alloy substrate. Y and Ti have been selected as thin-film materials because they form highly stable protective oxides layers. The Fe-12%Cr binary alloy has been selected because it is representative of ferritic steels that are widely used in nuclear systems. The absence of other alloying elements in this binary alloy would allow for a clearer examination of structures and compositions that evolve during high-temperature irradiations and oxidation treatments. The research is divided into four specific tasks: (1) sputter deposition of 500 nm thick films of Y and Ti on Fe-12%Cr alloy substrates, (2) ion irradiation of the film-substrate system with 2MeV protons to a dose of 2 dpa at temperatures of 300°C, 500°C, and 700°C, (3) oxidation of as-deposited and ion-irradiated samples in a controlled oxygen environment at 500°C and 700°C, (4) multi-scale computational modeling involving first- principle molecular dynamics (FPMD) ...
Date: January 9, 2014
Creator: Sridharan, Kumar & Zhang, Jinsuo
Partner: UNT Libraries Government Documents Department

The Suppression of Energy Discretization Errors in Multigroup Transport Calculations

Description: The Objective of this project is to develop, implement, and test new deterministric methods to solve, as efficiently as possible, multigroup neutron transport problems having an extremely large number of groups. Our approach was to (i) use the standard CMFD method to "coarsen" the space-angle grid, yielding a multigroup diffusion equation, and (ii) use a new multigrid-in-space-and-energy technique to efficiently solve the multigroup diffusion problem. The overall strategy of (i) how to coarsen the spatial and energy grids, and (ii) how to navigate through the various grids, has the goal of minimizing the overall computational effort. This approach yields not only the fine-grid solution, but also coarse-group flux-weighted cross sections that can be used for other related problems.
Date: June 17, 2013
Creator: Larsen, Edward
Partner: UNT Libraries Government Documents Department

Rapid Automated Dissolution and Analysis Techniques for Radionuclides in Recycle Process Streams

Description: The analysis of process samples for radionuclide content is an important part of current procedures for material balance and accountancy in the different process streams of a recycling plant. The destructive sample analysis techniques currently available necessitate a significant amount of time. It is therefore desirable to develop new sample analysis procedures that allow for a quick turnaround time and increased sample throughput with a minimum of deviation between samples. In particular new capabilities for rapid sample dissolution and radiochemical separation are required. Most of the radioanalytical techniques currently employed for sample analysis are based on manual laboratory procedures. Such procedures are time and labor intensive and not well suited for situations in which a rapid sample analysis is requires and/or large number of samples needed to be analyzed.
Date: July 18, 2013
Creator: Sudowe, Ralf
Partner: UNT Libraries Government Documents Department

Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term Irradiation at Elevated Temperature: Critical Experiments

Description: The in-service degradation of reactor core materials is related to underlying changes in the irradiated microstructure. During reactor operation, structural components and cladding experience displacement of atoms by collisions with neutrons at temperatures at which the radiation-induced defects are mobile, leading to microstructure evolution under irradiation that can degrade material properties. At the doses and temperatures relevant to fast reactor operation, the microstructure evolves by microchemistry changes due to radiation-induced segregation, dislocation loop formation and growth, radiation induced precipitation, destabilization of the existing precipitate structure, as well as the possibility for void formation and growth. These processes do not occur independently; rather, their evolution is highly interlinked. Radiation-induced segregation of Cr and existing chromium carbide coverage in irradiated alloy T91 track each other closely. The radiation-induced precipitation of Ni-Si precipitates and RIS of Ni and Si in alloys T91 and HCM12A are likely related. Neither the evolution of these processes nor their coupling is understood under the conditions required for materials performance in fast reactors (temperature range 300-600°C and doses to 200 dpa and beyond). Further, predictive modeling is not yet possible, as models for microstructure evolution must be developed along with experiments to characterize these key processes and provide tools for extrapolation. To extend the range of operation of nuclear fuel cladding and structural materials in advanced nuclear energy and transmutation systems to that required for the fast reactor, the irradiation-induced evolution of the microstructure, microchemistry, and the associated mechanical properties at relevant temperatures and doses must be understood. This project builds upon joint work at the proposing institutions, under a NERI-C program that is scheduled to end in September, to understand the effects of radiation on these important materials. The objective of this project is to conduct critical experiments to understand the evolution of microstructural and microchemical features ...
Date: December 20, 2013
Creator: Was, Gary; Jiao, Zhijie; Allen, Todd & Yang, Yong
Partner: UNT Libraries Government Documents Department

Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term and Elevated Temperature Irradiation: Modeling and Experimental Investigation

Description: The in-service degradation of reactor core materials is related to underlying changes in the irradiated microstructure. During reactor operation, structural components and cladding experience displacement of atoms by collisions with neutrons at temperatures at which the radiation-induced defects are mobile, leading to microstructure evolution under irradiation that can degrade material properties. At the doses and temperatures relevant to fast reactor operation, the microstructure evolves by dislocation loop formation and growth, microchemistry changes due to radiation-induced segregation, radiation-induced precipitation, destabilization of the existing precipitate structure, and in some cases, void formation and growth. These processes do not occur independently; rather, their evolution is highly interlinked. Radiationinduced segregation of Cr and existing chromium carbide coverage in irradiated alloy T91 track each other closely. The radiation-induced precipitation of Ni-Si precipitates and RIS of Ni and Si in alloys T91 and HCM12A are likely related. Neither the evolution of these processes nor their coupling is understood under the conditions required for materials performance in fast reactors (temperature range 300-600°C and doses beyond 200 dpa). Further, predictive modeling is not yet possible as models for microstructure evolution must be developed along with experiments to characterize these key processes and provide tools for extrapolation. To extend the range of operation of nuclear fuel cladding and structural materials in advanced nuclear energy and transmutation systems to that required for the fast reactor, the irradiation-induced evolution of the microstructure, microchemistry, and the associated mechanical properties at relevant temperatures and doses must be understood. Predictive modeling relies on an understanding of the physical processes and also on the development of microstructure and microchemical models to describe their evolution under irradiation. This project will focus on modeling microstructural and microchemical evolution of irradiated alloys by performing detailed modeling of such microstructure evolution processes coupled with well-designed in situ experiments that ...
Date: December 1, 2013
Creator: Wirth, Brian; Morgan, Dane; Kaoumi, Djamel & Motta, Arthur
Partner: UNT Libraries Government Documents Department

Quantification of UV-Visible and Laser Spectroscopic Techniques for Materials Accountability and Process Control

Description: Ultraviolet–visible spectroscopy (UV–Visible) and time-resolved laser fluorescence spectroscopy (TRLFS) optical techniques can permit on-line analysis of actinide elements in a solvent extraction process in real time. These techniques have been used for measuring actinide speciation and concentration under laboratory conditions and are easily adaptable to multiple sampling geometries, such as dip probes, fiber-optic sample cells, and flow-through cell geometries. To fully exploit these techniques, researchers must determine the fundamental speciation of target actinides and the resulting influence on spectroscopic properties. Detection limits, process conditions, and speciation of key actinide components can be established and utilized in a range of areas, particularly those related to materials accountability and process control. Through this project, researchers will develop tools and spectroscopic techniques to evaluate solution extraction conditions and concentrations of U, Pu, and Cm in extraction processes, addressing areas of process control and materials accountability. The team will evaluate UV– Visible and TRLFS for use in solvent extraction-based separations. Ongoing research is examining efficacy of UV-Visible spectroscopy to evaluate uranium and plutonium speciation under conditions found in the UREX process and using TRLFS to evaluate Cm speciation and concentration in the TALSPEAK process. A uranyl and plutonium nitrate UV–Visible spectroscopy study met with success, which supports the utility and continued exploration of spectroscopic methods for evaluation of actinide concentrations and solution conditions for other aspects of the UREX+ solvent extraction scheme. This project will ex examine U and Pu absorbance in TRUEX and TALSPEAK, perform detailed examination of Cm in TRUEX and TALSPEAK, study U laser fluorescence, and apply project data to contactors. The team will also determine peak ratios as a function of solution concentrations for the UV-Visible spectroscopy studies. The use of TRLFS to examine Cm and U will provide data to evaluate lifetime, peak location, and peak ratios (mainly for ...
Date: September 13, 2013
Creator: Czerwinski, Kenneth
Partner: UNT Libraries Government Documents Department

Investigation on the Core Bypass Flow in a Very High Temperature Reactor

Description: Uncertainties associated with the core bypass flow are some of the key issues that directly influence the coolant mass flow distribution and magnitude, and thus the operational core temperature profiles, in the very high-temperature reactor (VHTR). Designers will attempt to configure the core geometry so the core cooling flow rate magnitude and distribution conform to the design values. The objective of this project is to study the bypass flow both experimentally and computationally. Researchers will develop experimental data using state-of-the-art particle image velocimetry in a small test facility. The team will attempt to obtain full field temperature distribution using racks of thermocouples. The experimental data are intended to benchmark computational fluid dynamics (CFD) codes by providing detailed information. These experimental data are urgently needed for validation of the CFD codes. The following are the project tasks: • Construct a small-scale bench-top experiment to resemble the bypass flow between the graphite blocks, varying parameters to address their impact on bypass flow. Wall roughness of the graphite block walls, spacing between the blocks, and temperature of the blocks are some of the parameters to be tested. • Perform CFD to evaluate pre- and post-test calculations and turbulence models, including sensitivity studies to achieve high accuracy. • Develop the state-of-the art large eddy simulation (LES) using appropriate subgrid modeling. • Develop models to be used in systems thermal hydraulics codes to account and estimate the bypass flows. These computer programs include, among others, RELAP3D, MELCOR, GAMMA, and GAS-NET. Actual core bypass flow rate may vary considerably from the design value. Although the uncertainty of the bypass flow rate is not known, some sources have stated that the bypass flow rates in the Fort St. Vrain reactor were between 8 and 25 percent of the total reactor mass flow rate. If bypass flow rates ...
Date: October 22, 2013
Creator: Hassan, Yassin
Partner: UNT Libraries Government Documents Department

Consistent Multigroup Theory Enabling Accurate Course-Group Simulation of Gen IV Reactors

Description: The objective of this proposal is the development of a consistent multi-group theory that accurately accounts for the energy-angle coupling associated with collapsed-group cross sections. This will allow for coarse-group transport and diffusion theory calculations that exhibit continuous energy accuracy and implicitly treat cross- section resonances. This is of particular importance when considering the highly heterogeneous and optically thin reactor designs within the Next Generation Nuclear Plant (NGNP) framework. In such reactors, ignoring the influence of anisotropy in the angular flux on the collapsed cross section, especially at the interface between core and reflector near which control rods are located, results in inaccurate estimates of the rod worth, a serious safety concern. The scope of this project will include the development and verification of a new multi-group theory enabling high-fidelity transport and diffusion calculations in coarse groups, as well as a methodology for the implementation of this method in existing codes. This will allow for a higher accuracy solution of reactor problems while using fewer groups and will reduce the computational expense. The proposed research represents a fundamental advancement in the understanding and improvement of multi- group theory for reactor analysis.
Date: November 29, 2013
Creator: Rahnema, Farzad; Haghighat, Alireza & Ougouag, Abderrafi
Partner: UNT Libraries Government Documents Department

Development of A Self Biased High Efficiency Solid-State Neutron Detector for MPACT Applications

Description: Neutron detection is an important aspect of materials protection, accounting, and control for transmutation (MPACT). Currently He-3 filled thermal neutron detectors are utilized in many applications; these detectors require high-voltage bias for operation, which complicates the system when multiple detectors are used. In addition, due to recent increase in homeland security activity and the nuclear renaissance, there is a shortage of He-3, and these detectors become more expensive. Instead, cheap solid-state detectors that can be mass produced like any other computer chips will be developed. The new detector does not require a bias for operation, has low gamma sensitivity, and a fast response. The detection system is based on a honeycomb-like silicon device, which is filled with B-10 as the neutron converter; while a silicon p-n diode (i.e., solar cell type device) formed on the thin silicon wall of the honeycomb structure detects the energetic charged particles emitted from the B-10 conversion layer. Such a detector has ~40% calculated thermal neutron detection efficiency with an overall detector thickness of about 200 ?m. Stacking of these devices allows over 90% thermal neutron detection efficiency. The goal of the proposed research is to develop a high-efficiency, low-noise, self-powered solid-state neutron detector system based on the promising results of the existing research program. A prototype of this solid-state neutron detector system with sufficient detector size (up to 8-inch diam., but still portable and inexpensive) and integrated with interface electronics (e.g., preamplifier) will be designed, fabricated, and tested as a coincidence counter for MPACT applications. All fabrications proposed are based on silicon-compatible processing; thus, an extremely cheap detector system could be massively produced like any other silicon chips. Such detectors will revolutionize current neutron detection systems by providing a solid-state alternative to traditional gas-based neutron detectors.
Date: September 3, 2013
Creator: Danon, Yaron; Bhat, Ishwara & Jian-Qiang Lu, James
Partner: UNT Libraries Government Documents Department

Corrosion in Supercritical carbon Dioxide: Materials, Environmental Purity, Surface Treatments, and Flow Issues

Description: The supercritical CO{sub 2} Brayton cycle is gaining importance for power conversion in the Generation IV fast reactor system because of its high conversion efficiencies. When used in conjunction with a sodium fast reactor, the supercritical CO{sub 2} cycle offers additional safety advantages by eliminating potential sodium-water interactions that may occur in a steam cycle. In power conversion systems for Generation IV fast reactors, supercritical CO{sub 2} temperatures could be in the range of 30°C to 650°C, depending on the specific component in the system. Materials corrosion primarily at high temperatures will be an important issue. Therefore, the corrosion performance limits for materials at various temperatures must be established. The proposed research will have four objectives centered on addressing corrosion issues in a high-temperature supercritical CO{sub 2} environment: Task 1: Evaluation of corrosion performance of candidate alloys in high-purity supercritical CO{sub 2}: The following alloys will be tested: Ferritic-martensitic Steels NF616 and HCM12A, austenitic alloys Incoloy 800H and 347 stainless steel, and two advanced concept alloys, AFA (alumina forming austenitic) steel and MA754. Supercritical CO{sub 2} testing will be performed at 450°C, 550°C, and 650°C at a pressure of 20 MPa, in a test facility that is already in place at the proposing university. High purity CO{sub 2} (99.9998%) will be used for these tests. Task 2: Investigation of the effects of CO, H{sub 2}O, and O{sub 2} impurities in supercritical CO{sub 2} on corrosion: Impurities that will inevitably present in the CO{sub 2} will play a critical role in dictating the extent of corrosion and corrosion mechanisms. These effects must be understood to identify the level of CO{sub 2} chemistry control needed to maintain sufficient levels of purity to manage corrosion. The individual effects of important impurities CO, H{sub 2}O, and O{sub 2} will be investigated by adding them ...
Date: December 10, 2013
Creator: Sridharan, Kumar & Anderson, Mark
Partner: UNT Libraries Government Documents Department