Advanced Liquid-Liquid Separations. DENISE SCHUH (University of Wisconsin Madison, WI 53706) JOANNA MCFARLANE (Oak Ridge National Laboratory, Oak Ridge, TN, 37831)

Centrifugal contactors have many distinct advantages compared with conventional technologies for liquid-liquid extraction operations. The small size and high processing throughput of centrifugal contactors facilitate process intensification, i.e., the minimization of process size and material inventories. These reductions significantly improve process economics. Three systems were analyzed; diethylene glycol dibutyl ether (dibutyl carbitol) with nitric acid, 1-butyl-3-methyl imidazolium bis(perfluoroethyl sulfonyl)imide (bmim BETI) with water, and bmim BETI with cyclohexane. The data obtained can be used to create an accurate model to predict optimum performance. The systems were examined using a laboratory-scale, commercially available 5-cm-diameter contactor. Multiple data points were collected by varying the speed of the rotor, the weir size, the pump frequency, the pump stroke, the ratio between the two pumps, and the concentration of the liquids, which provided valuable information as to when the system would fail. Many physical and chemical properties of the three systems have also been determined, such as: calculating the surface tension using a contact angle meter; using a ring tensiometer, another method to measure the surface tension; determining the dispersion number by agitation; determining the density using Archimedes Principle; and determining the viscosity using a Brookfield Rheometer. For the dibutyl carbitol and nitric acid system, using weaker acid than 0.1M is not feasible due to the extremely poor separation time. Poor separation time was also measured with 1M nitric acid because the dibutyl carbitol was found to be contaminated with an unidentified substance. This particular system was examined because it is used in the purification of noble and heavy metals (e.g. gold and uranium) by selective extraction from aqueous solutions. The ionic liquid systems are currently being tested to gain more knowledge about them with hope they may be used for commercial processes in the future.

Benchmarking of Monte Carlo Codes and Sensitivity Analysis of Physics models in MCNPX for High Energy Projectile-Target Interaction. JERRAD DEASON (Prairie View A&M University Prairie View, TX 77446) JEFFREY O. JOHNSON (Oak Ridge National Laboratory, Oak Ridge, TN, 37831)

Neutral and charged particle transport phenomena is described with the Boltzmann transport equation. The numerical solution of the Boltzmann transport equation is applicable to various fields of nuclear engineering and physics. One such numerical technique, the Monte Carlo method, provides stochastic solutions to the particle transport calculations utilizing detailed three-dimensional geometry. Recent implementation of heavy ion transport capabilities has advanced the applicability of these codes to space and high energy physics problems. MCNPX (Monte Carlo N-Particle eXtended) and PHITS (Particle and Heavy-Ion Transport System) are two such multi-purpose particle and heavy ion transport codes that implement the Monte Carlo method and are ideally suited for characterization of high energy charged particle interactions and heavy ion transport design studies. However, the application of these codes currently has limitations as they are relatively new and unverified, particularly in the energy regimes where charged particle interaction data is sparse or absent. In these regions codes rely on physic model calculations to compute interaction probabilities. The work performed here was to validate and benchmark these two transport codes against the quantitative data collected from a peer-reviewed journal article. The article reports secondary neutron spectra measurements at discrete angles from a target to the scintillator detectors. The report presents the comparison of measured values with the computational results for two targets (water and lead) at nine different angles (0, 6, 15, 30, 45, 60, 90, 120 and 150) with a specific energy distribution at each angle. Additionally, for the lead target, calculations were performed at two different incident alpha energies (710-MeV and 640-MeV) and compared to the measured data. The resulting data extracted from the codes, are graphed on a log-log plot to illustrate the difference between the measured and computational neutron spectra at the various angles. The simulated geometries for the water target were then subjected to several alterations in the computational structures and physic mode implementations, such as, removing the columbic barrier parameter, and changing the physics models from the program defaults (Bertini and ISABEL) to the CEM2K model. These additional calculations were performed to compare the sensitivity of the physics models on the neutron production spectra.

Case Study of Radioactive Modeling with Irregular Source Formations. ANDREW ZURAWSKI (Purdue University West Lafayette, IN 47906) JAMES TERRY (Oak Ridge National Laboratory, Oak Ridge, TN, 37831)

Dose rates were modeled and measured for 20-foot cargo containers loaded with drums containing hydrated thorium nitrate in two configurations. One configuration consisted of 38 85-gallon drums each containing 825 pounds of hydrated thorium nitrate; the other configuration consisted of 127 30-gallon drums each containing 200 lbs of hydrated thorium nitrate. The decrease of the dose rates with distance were calculated using both line source and point source models. The modeled decrease of the dose rate with distance was compared to the measured dose rate. To relate the modeled unshielded source with the measurements, an effective linear attenuation coefficient was computed. Despite a geometry that appears to be more line-like than spherically symmetric, the point source predictions for dose rates were found to be in much better agreement with the measurements than were the predictions using the line source model.

Cost Analysis of the Hydrogen Economy Using FLOW. SHANNON WROBLEWSKI (Tennessee Technological University Cookeville, TN 38505) JUAN J. FERRADA (Oak Ridge National Laboratory, Oak Ridge, TN, 37831)

As the need for alternative fuel and energy sources increases, efforts to analyze the cost and viability of these sources also increase. One of the cost analysis models the Department of Energy (DOE) is currently using is known as H2 analysis, or H2A. H2A estimates the cost of production, delivery, and storage of hydrogen, giving a cost in $/kg of hydrogen. The H2A model looks at several different scenarios, such as transporting hydrogen as a compressed gas through a pipeline, as a compressed gas by truck, or as a liquid by truck. Currently, these models are all calculated in Excel on different spreadsheets. Using, reverse engineering, a step by step process was developed by which each step of the H2A model could be programmed into a simulation software package known as FLOW, named for its ability to create flow sheets and created at Oak Ridge National Laboratory. Python, a process modeling scripting language, was the computer language used due to the fact that it is already embedded in FLOW. After a module was made, it was then checked to make sure it corresponded to the values found in the H2A calculations. Then the module could be used in a flow sheet to compare similar types of operations, such as the transportation methods mentioned above. This allows the user to more easily compare different methods used in the H2A analysis in several ways. First, FLOW is a graphical program and does not show intermediate calculations as in the spreadsheets used in H2A. Second, several methods of transportation, storage, or other components of the hydrogen economy can be compared on the same flow sheet. Finally, user inputs are easily manipulated as in Excel. This project is a small but important part in the DOE's analysis of the hydrogen economy. It allows quick, yet accurate use of H2A analysis spreadsheets in an easy to use graphical format, allowing this small section of the hydrogen economy to be modeled more effectively.

Detection of Illicit Uranium Masked by the Presence of. STEVEN SAAVEDRA (University of New Mexico Albuquerque, NM 87131) IAN GROSS (Oak Ridge National Laboratory, Oak Ridge, TN, 37831)

Identification of unknown radioactive sources is a necessary process. With the heightened level of interest in the prevention of illicit trafficking of radioactive materials, new methods for the detection and characterization are being developed and evaluated. These methods are focused on detection of Uranium in the presence of more active radioisotopes to reduce the probability that an adversary could import fissile Uranium, an improvised nuclear device (IND) or a stolen nuclear weapon masked by a legitimate radioactive material shipment. An experiment was performed to detect the presence of Uranium by gamma spectroscopy utilizing a high purity Germanium (HPGe) radiation detector. A radioactive sample containing Uranium and an unknown radionuclide was placed in close proximity to a HPGe radiation detector with a Beryllium window end cap. An energy spectrum was collected for analysis. Analysis of the gamma peaks in the spectrum indicated that the sample was Ho-166m. None of the traditional photo peaks from Uranium were identified in the spectrum as the Ho-166m peaks overwhelmed the spectral results. By a non-traditional method of characteristic x-ray identification, the presence of Uranium was verified by identifying its characteristic x-ray peaks in the acquired spectrum. The energies of these characteristic x-rays, as listed in the nuclear decay tables, are 94.3 keV, 98.4 keV, 110.4 keV and 114.4 keV. The Uranium gamma peaks could not be seen in the acquired spectrum due to the large difference in specific activity on the order of 6 orders of magnitude between Uranium and Ho-166m, 1.922 10-6 Ci/g and 1.8 Ci/g respectively. The presence of Uranium, and other low specific activity radionuclides in the presence of high specific activity isotopes, can be verified through the non-traditional means of characteristic x-ray identification. This method is limited in the fact that the isotope can neither be qualified nor quantified through this method. To determine other artifacts such as enrichment or isotopic makeup more active means, such as sampling or active interrogation, would need to be employed.

Evaluation of a 3D Design Information Verification System for Uses at Oak Ridge National Lab. CAROL DUDNEY (University of Tennessee Knoxville, TN 37916) MICHEAL WHITAKER (Oak Ridge National Laboratory, Oak Ridge, TN, 37831)

Through a joint research cooperative with the European Commission's Joint Research Center, Oak Ridge National Laboratory (ORNL) has been given access to the 3D Design Information Verification (3D-DIV) system. This is a laser scanning system that can determine if an object has been moved as little as five millimeters. ORNL will test the 3D-DIV system to evaluate the systems limitations, applicability to ORNL, and function within nuclear installations. Five different experiments will be conducted. Scans will be taken from several well marked spots and subsequent verification scans will be taken at various distances from the original positions to test the importance of the placement of the laser. The use of barcodes in conjunction with the 3D-DIV system will be examined to determine other possible field applications. The response of the system to highly reflective materials at different angles will be evaluated. The next experiment will deal with layers of piping similar to that in nuclear facilities. And finally, the ability of the system to scan through different mediums (e.g. thin glass, colored water) will be assessed. The results obtained from these tests will be presented.

Evaluation of Electromagnetic Pump Data and Examination of Theoretical Mechanisms for Performance. VERNON GUTHRIE (Clemson University Clemson, SC 29631) GRAYDON L. YODER, JR. (Oak Ridge National Laboratory, Oak Ridge, TN, 37831)

The removal of heat developed in a nuclear reactor requires a pumping system to circulate a fluid through the reactor and then to a heat exchanger. Liquid metals are advantageous for use as the working fluid for cooling nuclear reactors because they have very high thermal conductivities and pumps can be used without moving parts. Electromagnetic (EM) pumps are often used for pumping liquid metals because of the problems conventional pumps have with liquid metals. Since EM pumps have no moving parts, they require no seals or bearing systems. The Annular Linear Induction Pump (ALIP) is a type of EM pump which produces flow in an electrically conductive fluid by the interaction of the currents induced in the liquid metal with the magnetic field created by the stator windings. Analysis of pump data obtained from a test loop containing an ALIP pumping liquid mercury provides insight into evaluating and maximizing pump performance. The product of differential pressure across the pump and volumetric flow rate through the pump is the work put into the fluid. Comparing this work to the product of the pump's input voltage and current allows the determination of pump efficiency. While pump data revealed efficiency values under 1%, these results were consistent with those expected for mercury ALIP pumps. The maximum efficiency calculated from test data was .78%. This value occurred at a differential pressure of 114.4 kPa, a volumetric flow rate .616 l/s, and a pump input power of 8.98 kW. In comparison, the efficiency value was .99% when using the pump manufacturer's data at the nominal operating point of 253.3 kPa, .4 l/s, and 10.2 kW. This effort supports a larger project designed to develop a space reactor power system capable of providing spacecraft power and propulsion for as long as fifteen years.

Evaluation of Russian Seals and Development of Wireless Tracking Systems for Nuclear Safeguards. DAVID YOUNKIN (University of Tennessee Knoxville, TN 37996) CHRIS PICKETT (Oak Ridge National Laboratory, Oak Ridge, TN, 37831)

Increased concern for national security is effectuating the development of technologies to, among other things, safeguard nuclear materials. To achieve this goal, wireless tracking systems can be used to monitor the location of a container and tamper-indicating devices (TIDs) can be used to detect unauthorized access into the container. However, these tools must be tested to determine their performance capabilities and characteristics before they can be effectively implemented. Russian supplied wire loop seals (OPP-1M) are TIDs that are noted for their simplicity of design. They are assembled by creating a unique wire pattern between two pieces of glass and then registered in a reader that can then verify the wire pattern image associated with the seal. These seals were tested to quantify the rate of failure of the reader in the laboratory environment as well as applied to drums to simulate an actual operating environment. The rate of failure in the laboratory is 2 percent in contrast to 4 percent when read on the drums. The data was also analyzed to identify factors that may influence this rate of failure. Another set of Russian TIDs was tested. These fiber-optic based seals transmit radio frequencies to a base station connected to a computer that records the status of the seal. Also, a wireless tracking system using 802.11 wireless routers and active radio frequency tags was set up and preliminary tests done to determine signal strength at given intervals. This system would allow containers to be tracked as they enter and move around a facility. Testing these safeguards reveals their strengths and weaknesses which will lead to a better determination of the levels of nuclear safeguards required to thwart different threats and to provide safer transport and storage of nuclear materials in the future.

Hydrogen Cost Analysis Simulation in Flow. QUAN COHEN (Robeson Community College Lumberton, NC 28358) DR. JUAN FERRADA (Oak Ridge National Laboratory, Oak Ridge, TN, 37831)

The Department of Energy (DOE) has designed a model, known as H2 Analysis (H2A), which addresses the economic aspects that must be considered while attempting to transition hydrogen as a fuel source into society. H2A has estimated all costs of production, storage, and modes of delivery from the storage facility to the distribution site in its "scenarios." The scenarios contain data that represent the implementation of a hydrogen economy such as projected feedstock and utility prices and labor costs. Delivery methods such as pipeline, truck tube-trailer equipment operation, and maintenance are included in the simulation. This module, however, will focus on the associated delivery costs of gaseous and liquid hydrogen from its storage facility to its distribution site in a truck tube-trailer. The project objective is to create a user-friendly interface that will display the main variables associated with hydrogen transportation. The user has the option of using H2A Projections in the code's database or entering other values if there is a variance. Flow is the simulation process and modeling tool used for analysis, developed by Oak Ridge National Laboratory (ORNL) that will be used to produce the module. Python is the scripting programming language used to execute Flow spreadsheets. The data used for Flow's code was extracted from two of DOE's twenty spreadsheets. The sequence of calculation steps were coded in Flow by using the reverse engineering method, which determined how the equations in the DOE spreadsheet were calculated. DOE spreadsheets had to be examined to determine the various forms of data compiled and differentiate the user-inputs from the calculated variables. This module produced results consistent with those produced by DOE's H2A spreadsheets. It allows for flexible and friendlier interaction with the user. In conclusion, this is a significantly small portion of the DOE hydrogen economy project.

Hydrogen Economy Costs Model Simulations. (HyTrans). MARCUS JACKSON (South Carolina State University. Orangeburg, SC 29117) JUAN FERRADDA (Oak Ridge National Laboratory, Oak Ridge, TN, 37831)

There is a need to develop interfaces between the already existing models and the global model of the hydrogen economy. Many transportation models have been developed using current oil fuels and future hydrogen fuel. However, these existing models and analytical tools are difficult to integrate and execute separately for analyzing efficiently and effectively the transition towards a hydrogen economy. Using the graphically oriented simulation software package FLOW and the process modeling scripting language Python various interfaces were developed to simulate different aspects of using hydrogen based on existing computerized transportation models (HyTrans developed at ORNL). FLOW simulated hydrogen economy scenarios by developing process flow sheets and analyzing capital, operating and unit production costs of these scenarios. Each scenario consisted of a selected production process, one delivery mechanism, and one type of retail. Several simulations were developed and run to compare the various alternatives and determine solutions beneficial to the U.S. Department of Energy Hydrogen Program. Results show that these simulations are accurate and consistent with the HyTrans model. This project is a small portion of a much larger project being pursued by the U.S. Department of Energy's Hydrogen Program to develop an efficient transition plan for a future use of hydrogen. In conclusion, the results suggest that this goal is achievable but current studies must be evaluated in order to accurately model an effective hydrogen economy.

Neutron and Gamma Ray Multiplet Analysis for the Determination of Plutonium Mass. MELISSA CRAWFORD (University of Florida Gainesville, FL 32608) SARA POZZI (Oak Ridge National Laboratory, Oak Ridge, TN, 37831)

Nuclear safeguards methods are being developed at ORNL and elsewhere to nondestructively determine the mass of fissile material. Recent efforts have studied the multiplicity of neutrons and gamma rays emitted by spontaneous and induced fissions in fissile samples. The present work aims at developing a reliable method for determining the mass of the samples. The proposed analysis relies on a series of experiments that were conducted on plutonium oxide samples at the JRC in Ispra, Italy, using a GHz processor developed at ORNL. The detectors used were liquid scintillators, which are sensitive to fast neutrons and gamma rays. The setup consisted of two 2x2 detector arrays, placed opposite each other, with a plutonium oxide sample set equidistant from them, inside of an AT400 container. The measurement setup was simulated using the MCNP-PoliMi code, and multiplet analysis was performed on both the simulated and the measured data. The distribution of multiplets describes the number of times n detections occur following a single fission, where n includes neutron and gamma ray counts. This work shows that it is possible to determine a relationship between the multiplet distribution and the mass of Pu-240 in the sample. Our results show that the Pu-240 mass can be predicted to within 7.8% in the mass range of 50 to 2500g. This information, together with information on the sample's composition, allows us to determine the mass of Pu-239. This work has a bearing on nuclear nonproliferation programs and will lead to more sophisticated analyses that can be applied to homeland security issues.

Nuclear Data Extracted for Use in Burnup Credit Calculations. SUSAN WILLIAMS (Texas A&M University College Station, TX 77843) DON MUELLER (Oak Ridge National Laboratory, Oak Ridge, TN, 37831)

Burnup credit is a negative reactivity credit caused by a decrease in the amount of uranium-235 and the accumulation of other nuclides in nuclear reactor fuel while it has been used to generate power. Realistic bounding operating conditions can be used in fuel burnup modeling to support realistic conservative burnup credit calculations. Reactor operating data extracted, organized, and processed from documents from several commercial nuclear power plants offer valuable information for nuclear fuel burnup calculations. Axial burnup distributions, fuel temperatures, and moderator densities were among some of the information extracted for reactors such as Crystal River, Sequoyah, Three Mile Island, and Quad Cities. In addition, soluble boron concentrations were obtained for pressurized water reactors. This and additional data will be used to more accurately model nuclear fuel burnup. This work is part of a larger project that's main goal is to safely and economically transport commercial spent nuclear fuel in casks.

Numerical Analysis of Benchmark data using MCNPX. ERIC WRIGHT (Prairie View A&M University Prairie View, TX 77446) JEFFREY O. JOHNSON (Oak Ridge National Laboratory, Oak Ridge, TN, 37831)

Developments in the radiation transport codes for high energy radiation transport have triggered the applicability of these codes to new areas of research within the high energy physics and the NASA communities. These codes are now being utilized to analyze secondary particle production from Galactic Cosmic Ray (GCR) and Solar Particle Event (SPE) interactions in space environments. Similarly, they are also applicable to the analyses associated with target selection and design for advanced high energy particle accelerators and associated shielding. MCNPX (Monte Carlo N-Particle eXtended), an extended version of the industry standard Monte Carlo N-Particle transport code (MCNP), is being evaluated for such applications. MCNPX provides a complete 3D simulation of the problem geometry. It also provides the ability to utilize nuclear cross section libraries, when available, and physics models for particle types and energies where tabular nuclear data are not available. The results from MCNPX calculations, performed to benchmark neutron production from high energy source (alpha) - target interactions, were compared to measured values reported in a peer-reviewed journal article. This work was performed to better understand the limitations of the MCNPX physics models and to validate the code's applicability to heavy ion accelerator target design. The analysis was performed for a focused 710 MeV alpha ion beam on four different target materials: Water, Carbon, Lead and Stainless Steel (SS). The computational results are compared with the measured data at nine different angles; seven forward angles (0, 6, 15, 30, 45, 60, and 90) and two back angles (120 and 150) with a specific energy distribution at each angle. The source-target-detector geometry was modeled explicitly with all the details that were documented in the journal article. The calculations were performed to attain 10% uncertainty or better. In most cases sixty million source particles were simulated in a parallel mode (Parallel Virtual Machine, PVM) on a 24 node Linux Cluster, "cpile," within the Nuclear Science and Technology Division (NSTD) of Oak Ridge National Laboratory (ORNL). This work supports the benchmark efforts of the Radiation Shielding Group of NSTD for applying MCNPX to target station design for particle accelerators such as the Rare Isotope Accelerator (RIA) and Spallation Neutron Source (SNS).

The Density of Actinide Solutions at Varying Concentrations and Temperatures. RACHEAL MARTINDALE (Purdue University West Lafayette, IN 47907) CHARLES WEBER (Oak Ridge National Laboratory, Oak Ridge, TN, 37831)

In order to build more effective models, nuclear criticality safety research has recently found the need to obtain additional density data for aqueous solutions of various actinide compounds as well as hydrofluoric acid. Information was collected from internal Oak Ridge National Laboratory (ORNL) sources in addition to open literature. Internal sources include logbooks belonging to ORNL researchers beginning in the 1950's in the areas of criticality research and chemical science and technology research, the annual progress reports of the Neutron Physics Division of ORNL from the same time period, and other ORNL internal research reports. Open literature searched includes journals and compilations published in the chemical and nuclear sciences. Data regarding aqueous solutions of uranyl fluoride (UO2F2), uranyl sulfate (UO2SO4), uranyl chloride (UO2Cl2), thorium chloride (ThCl4), thorium nitrate (Th(NO3)4), and hydrofluoric acid (HF) was found and compiled. This data was then evaluated statistically and analyzed for error in preparation for input into a small experimental program.

Validation of SCALE (TRITON) Isotopic Predictions for Light Water Reactor Spent Fuel. DANIEL GILL (Pennsylvania State University University Park, PA 16802) STEVE BOWMAN (Oak Ridge National Laboratory, Oak Ridge, TN, 37831)

The ability to accurately predict the nuclide composition of spent fuel samples over the course of time is important in a wide variety of applications. These applications include, but are not limited to, the design, licensing and operation of radioactive waste transport systems, interim waste storage, and a permanent waste repository. The isotopic depletion capabilities of the new SCALE (Standardized Computer Analyses for Licensing Evaluation) control module TRITON, coupled with ORIGEN-S, were evaluated using spent fuel assays from several commercial light water reactors. The type of reactor analyzed was the pressurized-water-reactor (PWR) with both standard and mixed-oxide fuel (MOX) assemblies. Calculations were performed using the functional modules NEWT and KENO-VI. NEWT is a two-dimensional, arbitrary geometry, discrete ordinates transport code and KENO-VI is a multigroup, Monte Carlo transport code capable of handling complex three-dimensional geometries. To validate the codes and data used in depletion calculations, numerical predictions were compared with experimental measurements for a total of 21 samples taken from the Calvert Cliffs, Obrigheim, and San Onofre PWRs. Similar comparisons have previously been performed at the Oak Ridge National Laboratory (ORNL) for the one-dimensional SAS2H control module. The SAS2H, TRITON/KENO-VI, and TRITON/NEWT results were compared for corresponding samples. All analyses showed that TRITON/KENO-VI and TRITON/NEWT produced typically similar results, leading to the conclusion that the transport/depletion models are performing correctly. For the San Onofre MOX assemblies, TRITON obtained depletion results nearly the same as or better than SAS2H for all nuclides considered. The average difference between calculated and measured values was below 1 % for 235Uand around 5 %for 239Pu. The results for the Calvert Cliffs and Obrigheim PWRs were in most cases also comparable to previous SAS2H results. The calculations performed in this validation study demonstrate that the depletion capabilities of TRITON accurately model spent fuel depletion and decay and are similar to those of SAS2H, which previous ORNL studies have concluded qualifies as a basic tool for predicting isotopic compositions of spent fuel.