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Environmental Assessment Methodology for the Nuclear Fuel Cycle

Description: This report describes the methodology for determining where environmental control technology is required for the nuclear fuel cycle. The methodology addresses routine emission of chemical and radioactive effluents, and applies to mining, milling, conversion, enrichment, fuel fabrication, reactors (LWR and BWR) and fuel reprocessing. Chemical and radioactive effluents are evaluated independently. Radioactive effluents are evaluated on the basis of maximum exposed individual dose and population dose calculations for a 1-year emission period and a 50-year commitment. Sources of radionuclides for each facility are then listed according to their relative contribution to the total calculated dose. Effluent, ambient and toxicology standards are used to evaluate the effect of chemical effluents. First, each chemical and source configuration is determined. Sources are tagged if they exceed existirrg standards. The combined effect of all chemicals is assessed for each facility. If the additive effects are unacceptable, then additional control technology is recommended. Finally, sources and their chemicals at each facility are ranked according to their relative contribution to the ambient pollution level. This ranking identifies those sources most in need of environmental control.
Date: July 1, 1977
Creator: Brenchley, D. L.; Soldat, J. K.; McNeese, J. A. & Watson, E. C.
Partner: UNT Libraries Government Documents Department

Multilaboratory analytical quality control for the hydrochemical and stream sediment reconnaissance

Description: For the first time data received from LLL has been incorporated in the quality assurance report. LASL has indicated that their results on the water standard B2 have been consistently low. They suspect that this may be resulting from uranium plating out in the polyethylene containers. Such an observation has not been made by either ORGDP, SRL or LLL. To evaluate these suspect observations LASL has been sent two sets of standards, one set contained in teflon and the second in the usual polyethylene containers. LASL results for June will be carefully evaluated.
Date: May 31, 1978
Creator: D'Silva, A. P.; Haas, W. J., Jr. & Floyd, M. A.
Partner: UNT Libraries Government Documents Department

Radkowsky Thorium Fuel Project

Description: In the early/mid 1990’s Prof. Alvin Radkowsky, former chief scientist of the U.S. Naval Reactors program, proposed an alternate fuel concept employing thorium-based fuel for use in existing/next generation pressurized water reactors (PWRs). The concept was based on the use of a 'seed-blanket-unit' (SBU) that was a one-for-one replacement for a standard PWR assembly with a uranium-based central 'driver' zone, surrounded by a 'blanket' zone containing uranium and thorium. Therefore, the SBU could be retrofit without significant modifications into existing/next generation PWRs. The objective was to improve the proliferation and waste characteristics of the current once-through fuel cycle. The objective of a series of projects funded by the Initiatives for Proliferation Prevention program of the U.S. Department of Energy (DOE-IPP) - BNL-T2-0074,a,b-RU 'Radkowsky Thorium Fuel (RTF) Concept' - was to explore the characteristics and potential of this concept. The work was performed under several BNL CRADAs (BNL-C-96-02 and BNL-C-98-15) with the Radkowsky Thorium Power Corp./Thorium Power Inc. and utilized the technical and experimental capabilities in the Former Soviet Union (FSU) to explore the potential of this concept for implementation in Russian pressurized water reactors (VVERs), and where possible, also generate data that could be used for design and licensing of the concept for Western PWRs. The Project in Russia was managed by the Russian Research Center-'Kurchatov Institute'(RRC-KI), and included several institutes (e.g., PJSC 'Electrostal', NPO 'LUCH' (Podolsk), RIINM (Bochvar Institute), GAN RF (Gosatomnadzor), Kalininskaja NPP (VVER-1000)), and consisted of the following phases: Phase-1 ($550K/$275K to Russia): The objective was to perform an initial review of all aspects of the concept (design, performance, safety, implementation issues, cost, etc.) to confirm feasibility/viability and identify any “show-stoppers”; Phase-2 ($600K/$300K to Russia): Continued the activities initiated under Phase-1 with a focus on expanded design and safety analyses, and to address fuel fabrication and testing ...
Date: December 31, 2006
Creator: Todosow, Michael
Partner: UNT Libraries Government Documents Department

Interpretation of Tracer Surface Diffusion Experiments on UO{Sub 2} Roles of Gas and Solid Transport Processes

Description: The spreading of a tracer from an enriched needle source which contacts the surface of a depleted pellet sink is analyzed rigorously. It is shown that volume diffusion in both the needle and the pellet need to be considered because only by this process is sufficient radioactivity accumulated for measurement after the anneal. Parasitic gas phase processes are of two types-evaporative loss of solid if a flowing gas is used, or molecular diffusion from enriched portions of the surface to depleted zones if the couple is in a closed vessel with a stagnant gas. A complete numerical solution including surface diffusion, solid diffusion, evaporative loss and contact resistance is applied to the UO{sub 2} tracer study of Marlowe and Kazanoff at 1915° C. Based upon UO{sub 2} evaporation experiments, the analysis shows that the evaporative loss effect is not important in these experiments. The UO{sub 2} surface diffusion coefficient deduced from analysis of these data is 0.2{+-} 0.1 cm{sup 2)/s at 1915{degrees}C., which is 10{sup 4} times larger than that predicted by extrapolation of values obtained by mass transfer techniques.
Date: June 1, 1980
Creator: Olander, D. R.
Partner: UNT Libraries Government Documents Department

Quantify Water Extraction by TBP/Dodecane via Molecular Dynamics Simulations

Description: The purpose of this project is to quantify the interfacial transport of water into the most prevalent nuclear reprocessing solvent extractant mixture, namely tri-butyl- phosphate (TBP) and dodecane, via massively parallel molecular dynamics simulations on the most powerful machines available for open research. Specifically, we will accomplish this objective by evolving the water/TBP/dodecane system up to 1 ms elapsed time, and validate the simulation results by direct comparison with experimentally measured water solubility in the organic phase. The significance of this effort is to demonstrate for the first time that the combination of emerging simulation tools and state-of-the-art supercomputers can provide quantitative information on par to experimental measurements for solvent extraction systems of relevance to the nuclear fuel cycle. Results: Initially, the isolated single component, and single phase systems were studied followed by the two-phase, multicomponent counterpart. Specifically, the systems we studied were: pure TBP; pure n-dodecane; TBP/n-dodecane mixture; and the complete extraction system: water-TBP/n-dodecane two phase system to gain deep insight into the water extraction process. We have completely achieved our goal of simulating the molecular extraction of water molecules into the TBP/n-dodecane mixture up to the saturation point, and obtained favorable comparison with experimental data. Many insights into fundamental molecular level processes and physics were obtained from the process. Most importantly, we found that the dipole moment of the extracting agent is crucially important in affecting the interface roughness and the extraction rate of water molecules into the organic phase. In addition, we have identified shortcomings in the existing OPLS-AA force field potential for long-chain alkanes. The significance of this force field is that it is supposed to be optimized for molecular liquid simulations. We found that it failed for dodecane and/or longer chains for this particular solvent extraction application. We have proposed a simple way to circumvent ...
Date: May 16, 2013
Creator: Khomami, Bamin; Cui, Shengting; de Almeida, Valmor F. & Felker, Kevin
Partner: UNT Libraries Government Documents Department

Reactivity Initiated Accident Test Series RIA Scoping Test Experiment Predictions

Description: The Reactivity 'Initiated Accident (RIA) test series to be conducted in the Power Burst Facility (PBF) has been designed.to determine fuel failure thresholds, modes, and consequences as a function of energy deposition, irradiation history, and fuel design. The RIA Scoping Test will be comprised of five single unirradiated rod sub-tests. The first rod will be subjected to a series of transient power bursts of increasing energy release to determine the energy deposition at cladding failure. The second and third rods will be subjected to energy depositions near that which caused failure of the first rod, to further define the failure threshold. Rods four and five will be subjected to large radially averaged energy depositions, 1990 and 2510 J/g respectively, to investigate facility safety concerns. Several analyses were performed to predict test fuel rod and system behavior during the five RIA Scoping Test phases. A reactor physics analysis was performed to obtain the relationship between test fuel rod and reactor core energy during a power transient. The calculations were made with the RAFFLE computer code. The thermal-hydraulic behavior of the test rod coolant was investigated for pellet surface energy depositions of 900, 1125, and 1350 J/g for the first three phases of the Scoping Test. The RELAP4 computer code was used for these thermal-hydraulic analyses. The results of the RELAP4 calculations provided input to the FRAP-T4 computer code for three fuel rod behavior analyses at pellet surface energy depositions of 815, 1020, and 1225 J/g. A cladding embrittlement analysis, using the results of the FRAP-T4 calculations as input, was made to investigate the cladding oxidation mode of rod failure for the lower energy phases. BUILD5 was the analytical tool used in this investigation. Finally, the pressure pulses generated as a result of failure of the test fuel rods in the final ...
Date: June 1978
Creator: Semken, R. S.; Eaton, A. M.; Smith, R. H. & Resch, S. C.
Partner: UNT Libraries Government Documents Department

Reactivity Initiated Accident Test Series RIA Scoping Test Quick Look Report

Description: The Reactivity Initiated Accident Scoping Test (RIA-ST) was successfully completed August 30, 1978. The test was introductory to the RIA Series 1 tests and was designed to investigate and resolve several anticipated problem areas prior to performance of the first test of the series, Test RIA 1-1. The RIA Scoping Test, as performed, consisted of four separate single-rod experiment phases. The first three phases were performed with shrouded fuel rods of 5.8 wt.% enrichment. They were subjected to power bursts resulting in total fuel surface energies ranging from 205 to 261 cal/q at the axial peak elevation. The fourth phase consisted of a 20 wt.% enriched, shrouded fuel rod which was subjected to a power hurst that deposited a total radially averaged energy of 527 cal/g. The primary objectives of the Scoping Test were defined as follows: (1) Determine the applicability of extrapolating low-power steady state calorimetric measurements and self-powered neutron detector (SPND) output to determine fuel rod energy depositions during a power burst. (2) Determine the enerqy deposition failure threshold for unirradiated fuel rods at BWR hot-startup coolant conditions. (3) Determine the magnitudes of oossible pressure pulses resulting from rod failure. (4) Determine the sensitivity of the test instrumentation to high transient radiation exposures. In general, the energy deposition values for the Scoping Test derived from the SPND output were 25% higher than those obtained from the core ion chamber data. Determining which values are correct will require radiochemical analysis of the fuel rods which will take several months. At present, it apoears that the SPND derived energies are in error because of excellent agreement between the calculated and measured power calibration results and the agreement between the predicted failure threshold and that seen using the core ion chamber derived energies. Meeting the second objective was accomplished during the ...
Date: September 1978
Creator: Martinson, Z. R.; Semken, R. S.; Inabe, T.; Smith, R. H.; Cook, T. F. & Appelhans, A. D.
Partner: UNT Libraries Government Documents Department

Technical Support for Improving the Licensing Regulatory Base for Selected Facilities Associated with the Front End of the Fuel Cycle

Description: Pacific Northwest Laboratory (PNL) was asked by the NRC Office of Nuclear Material Safety and Safeguards (NMSS) to determine the adequacy of its health, safety and environmental regulatory base as a guide to applicants for licenses to operate UF{sub 6} conversion facilities and fuel fabrication plants. The regulatory base was defined as the body of documented requirements and guidance to licensees, including laws passed by Congress, Federal Regulations developed by the NRC to implement the laws, license conditions added to each license to deal with special requirements for that specific license, and Regulatory Guides. The study concentrated on the renewal licensing accomplished in the last few years at five typical facilities, and included analyses of licensing documents and interviews with individuals involved with different aspects of the licensing process. Those interviewed included NMSS staff, Inspection and Enforcement (IE) officials, and selected licensees. From the results of the analyses and interviews, the PNL study team concludes that the regulatory base is adequate but should be codified for greater visibility. PNL recommends that NMSS clarify distinctions among legal requirements of the licensee, acceptance criteria employed by NMSS, and guidance used by all. In particular, a prelicensing conference among NMSS, IE and each licensee would be a practical means of setting license conditions acceptable to all parties.
Date: April 1, 1982
Creator: Clark, R. G.; Schreiber, R. E.; Jamison, J. D.; Davenport, L. C. & Brite, D. W.
Partner: UNT Libraries Government Documents Department

Enterprise SRS: Leveraging Ongoing Operations To Advance Nuclear Fuel Cycles Research And Development Programs

Description: The Savannah River Site (SRS) is repurposing its vast array of assets to solve future national issues regarding environmental stewardship, national security, and clean energy. The vehicle for this transformation is Enterprise SRS which presents a new, radical view of SRS as a united endeavor for ''all things nuclear'' as opposed to a group of distinct and separate entities with individual missions and organizations. Key among the Enterprise SRS strategic initiatives is the integration of research into facilities in conjunction with on-going missions to provide researchers from other national laboratories, academic institutions, and commercial entities the opportunity to demonstrate their technologies in a relevant environment and scale prior to deployment. To manage that integration of research demonstrations into site facilities, The Department of Energy, Savannah River Operations Office, Savannah River Nuclear Solutions, the Savannah River National Laboratory (SRNL) have established a center for applied nuclear materials processing and engineering research (hereafter referred to as the Center). The key proposition of this initiative is to bridge the gap between promising transformational nuclear fuel cycle processing discoveries and large commercial-scale-technology deployment by leveraging SRS assets as facilities for those critical engineering-scale demonstrations necessary to assure the successful deployment of new technologies. The Center will coordinate the demonstration of R&D technologies and serve as the interface between the engineering-scale demonstration and the R&D programs, essentially providing cradle-to-grave support to the research team during the demonstration. While the initial focus of the Center will be on the effective use of SRS assets for these demonstrations, the Center also will work with research teams to identify opportunities to perform research demonstrations at other facilities. Unique to this approach is the fact that these SRS assets will continue to accomplish DOE's critical nuclear material missions (e.g., processing in H-Canyon and plutonium storage in K-Area). Thus, the ...
Date: July 3, 2013
Creator: Murray, Alice M.; Marra, John E.; Wilmarth, William R.; Mcguire, Patrick W. & Wheeler, Vickie B.
Partner: UNT Libraries Government Documents Department


Description: The Savannah River Site's (SRS) H Canyon Facility is the only large scale, heavily shielded, nuclear chemical separations plant still in operation in the U.S. The facility's operations historically recovered uranium-235 (U-235) and neptunium-237 (Np-237) from aluminum-clad, enriched-uranium fuel tubes from Site nuclear reactors and other domestic and foreign research reactors. Today the facility, in conjunction with HB Line, is working to provide the initial feed material to the Mixed Oxide Facility also located on SRS. Many additional campaigns are also in the planning process. Furthermore, the facility has started to integrate collaborative research and development (R&D) projects into its schedule. H Canyon can serve as the appropriate testing location for many technologies focused on monitoring the back end of the fuel cycle, due to the nature of the facility and continued operation. H Canyon, in collaboration with the Savannah River National Laboratory (SRNL), has been working with several groups in the DOE complex to conduct testing demonstrations of novel technologies at the facility. The purpose of conducting these demonstrations at H Canyon will be to demonstrate the capabilities of the emerging technologies in an operational environment. This paper will summarize R&D testing activities currently taking place in H Canyon and discuss the possibilities for future collaborations.
Date: July 9, 2013
Creator: Sexton, L. & Fuller, Kenneth
Partner: UNT Libraries Government Documents Department


Description: Reactive Gas Recycling (RGR) technology development has been initiated at Savannah River National Laboratory (SRNL), with a stretch-goal to develop a fully dry recycling technology for Used Nuclear Fuel (UNF). This approach is attractive due to the potential of targeted gas-phase treatment steps to reduce footprint and secondary waste volumes associated with separations relying primarily on traditional technologies, so long as the fluorinators employed in the reaction are recycled for use in the reactors or are optimized for conversion of fluorinator reactant. The developed fluorination via SF{sub 6}, similar to the case for other fluorinators such as NF{sub 3}, can be used to address multiple fuel forms and downstream cycles including continued processing for LWR via fluorination or incorporation into a aqueous process (e.g. modified FLUOREX) or for subsequent pyro treatment to be used in advanced gas reactor designs such metal- or gas-cooled reactors. This report details the most recent experimental results on the reaction of SF{sub 6} with various fission product surrogate materials in the form of oxides and metals, including uranium oxides using a high-temperature DTA apparatus capable of temperatures in excess of 1000{deg}C . The experimental results indicate that the majority of the fission products form stable solid fluorides and sulfides, while a subset of the fission products form volatile fluorides such as molybdenum fluoride and niobium fluoride, as predicted thermodynamically. Additional kinetic analysis has been performed on additional fission products. A key result is the verification that SF{sub 6} requires high temperatures for direct fluorination and subsequent volatilization of uranium oxides to UF{sub 6}, and thus is well positioned as a head-end treatment for other separations technologies, such as the volatilization of uranium oxide by NF{sub 3} as reported by colleagues at PNNL, advanced pyrochemical separations or traditional full recycle approaches. Based on current results of ...
Date: September 25, 2012
Creator: Gray, J.; Torres, R.; Korinko, P.; Martinez-Rodriguez, M.; Becnel, J.; Garcia-Diaz, B. et al.
Partner: UNT Libraries Government Documents Department

From the Lab to the real world : sources of error in UF {sub 6} gas enrichment monitoring

Description: Safeguarding uranium enrichment facilities is a serious concern for the International Atomic Energy Agency (IAEA). Safeguards methods have changed over the years, most recently switching to an improved safeguards model that calls for new technologies to help keep up with the increasing size and complexity of today’s gas centrifuge enrichment plants (GCEPs). One of the primary goals of the IAEA is to detect the production of uranium at levels greater than those an enrichment facility may have declared. In order to accomplish this goal, new enrichment monitors need to be as accurate as possible. This dissertation will look at the Advanced Enrichment Monitor (AEM), a new enrichment monitor designed at Los Alamos National Laboratory. Specifically explored are various factors that could potentially contribute to errors in a final enrichment determination delivered by the AEM. There are many factors that can cause errors in the determination of uranium hexafluoride (UF{sub 6}) gas enrichment, especially during the period when the enrichment is being measured in an operating GCEP. To measure enrichment using the AEM, a passive 186-keV (kiloelectronvolt) measurement is used to determine the {sup 235}U content in the gas, and a transmission measurement or a gas pressure reading is used to determine the total uranium content. A transmission spectrum is generated using an x-ray tube and a “notch” filter. In this dissertation, changes that could occur in the detection efficiency and the transmission errors that could result from variations in pipe-wall thickness will be explored. Additional factors that could contribute to errors in enrichment measurement will also be examined, including changes in the gas pressure, ambient and UF{sub 6} temperature, instrumental errors, and the effects of uranium deposits on the inside of the pipe walls will be considered. The sensitivity of the enrichment calculation to these various parameters will then be ...
Date: March 1, 2012
Creator: Lombardi, Marcie L.
Partner: UNT Libraries Government Documents Department

Fundamental Studies of Irradiation-Induced Defect Formation and Fission Product Dynamics in Oxide Fuels

Description: The objective of this research program is to address major nuclear fuels performance issues for the design and use of oxide-type fuels in the current and advanced nuclear reactor applications. Fuel performance is a major issue for extending fuel burn-up which has the added advantage of reducing the used fuel waste stream. It will also be a significant issue with respect to developing advanced fuel cycle processes where it may be possible to incorporate minor actinides in various fuel forms so that they can be 'burned' rather than join the used fuel waste stream. The potential to fission or transmute minor actinides and certain long-lived fission product isotopes would transform the high level waste storage strategy by removing the need to consider fuel storage on the millennium time scale.
Date: December 19, 2012
Creator: Stubbins, James
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


Description: This report presents the results of a study to compare and evaluate the safety and economics of spent fuel shipments by special trains relative to shipments by conventional freight trains. The safety of special trains was investigated by analyzing data from the Federal Railroad Administration for freight train accidents occurring during the period 1972-1974. The economic analysis compared the cost of spent fuel shipments by special trains and by conventional freight trains.
Date: December 1, 1977
Creator: Loscutoff, W. V.; Murphy, E. S.; Clark, L. L.; McKee, R. W. & Hall, R. J.
Partner: UNT Libraries Government Documents Department

Solubility Classification of Airborne Uranium Products from LWR-Fuel Plants

Description: Airborne dust samples were obtained from various locations within plants manufacturing fuel elements for light-water reactors, and the dissolution rates of uranium from these samples into simulated lung fluid at 37°C were measured. These measurements were used to classify the solubilities of the samples in terms of the lung clearance model proposed by the International Commission on Radiological Protection. Similar evaluations were performed for samples of pure uranium compounds expected as components in plant dust. The variation in solubility classifications of dust encountered along the fuel production lines is described and correlated with the process chemistry and the solubility classifications of the pure uranium compounds.
Date: August 1, 1980
Creator: kalkwarf, D. R.
Partner: UNT Libraries Government Documents Department

Simulations of the Thermodynamic and Diffusion Properties of Actinide Oxide Fuel Materials

Description: Spent nuclear fuel from commercial reactors is comprised of 95-99 percent UO{sub 2} and 1-5 percent fission products and transuranic elements. Certain actinides and fission products are of particular interest in terms of fuel stability, which affects reprocessing and waste materials. The transuranics found in spent nuclear fuels are Np, Pu, Am, and Cm, some of which have long half- lives (e.g., 2.1 million years for {sup 237}Np). These actinides can be separated and recycled into new fuel matrices, thereby reducing the nuclear waste inventory. Oxides of these actinides are isostructural with UO{sub 2}, and are expected to form solid solutions. This project will use computational techniques to conduct a comprehensive study on thermodynamic properties of actinide-oxide solid solutions. The goals of this project are to: Determine the temperature-dependent mixing properties of actinide-oxide fuels; Validate computational methods by comparing results with experimental results; Expand research scope to complex (ternary and quaternary) mixed actinide oxide fuels. After deriving phase diagrams and the stability of solid solutions as a function of temperature and pressure, the project team will determine whether potential phase separations or ordered phases can actually occur by studying diffusion of cations and the kinetics of potential phase separations or ordered phases. In addition, the team will investigate the diffusion of fission product gases that can also have a significant influence on fuel stability. Once the system has been established for binary solid solutions of Th, U, Np, and Pu oxides, the methodology can be quickly applied to new compositions that apply to ternaries and quaternaries, higher actinides (Am, Cm), burnable poisons (B, Gd, Hf), and fission products (Cs, Sr, Tc) to improve reactivity.
Date: April 16, 2013
Creator: Becker, Udo
Partner: UNT Libraries Government Documents Department


Description: This report presents a summary of testing the affinity of titanosilicates (TSP), germanium-substituted titanosilicates (Ge-TSP) and multiwall carbon nanotubes (MWCNT) for lanthanide ions in dilute nitric acid solution. The K-TSP ion exchanger exhibited the highest affinity for lanthanides in dilute nitric acid solutions. The Ge-TSP ion exchanger shows promise as a material with high affinity, but additional tests are needed to confirm the preliminary results. The MWCNT exhibited much lower affinities than the K-TSP in dilute nitric acid solutions. However, the MWCNT are much more chemically stable to concentrated nitric acid solutions and, therefore, may candidates for ion exchange in more concentrated nitric acid solutions. This technical report serves as the deliverable documenting completion of the FY13 research milestone, M4FT-13SR0303061 – measure actinide and lanthanide distribution values in nitric acid solutions with sodium and potassium titanosilicate materials.
Date: April 24, 2013
Creator: Alsobrook, A. & Hobbs, D.
Partner: UNT Libraries Government Documents Department

FUEL PERFORMANCE IMPROVEMENT PROGRAM Thermal Conductivity of Sphere-Pac Fuel

Description: Progress in understanding the thermal conductivity of sphere-pac fuel beds has been made both at Oregon State University and Exxon Nuclear Company supported by the Fuel Performance Improvement Program (FPIP). FPIP is sponsored by the U. S. Department of Energy and is being performed by Consumers Power Company, Exxon Nuclear Company, and Pacific Northwest Laboratory. The purpose of the program is to test and demonstrate improved li9ht water reactor fuel concepts that are more resistant to failure from pellet-cladding interaction during power increases than standard pellet fuel.
Date: July 1, 1981
Creator: Ades,, M. J.
Partner: UNT Libraries Government Documents Department


Description: The objectives of the Fuel Performance Improvement Program (FPIP) are to identify and demonstrate fuel concepts with improved performance and to provide the supportive technical bases for developing commercial fuel designs that are capable of achieving hiqh burnup for better utilization of uranium.
Date: July 1, 1978
Creator: Crouthamel, CE
Partner: UNT Libraries Government Documents Department