Nanomaterials and nanotechnology methods have been an integral part of international research over the past decade. Because many traditional water treatment technologies (e.g. membrane filtration, biofouling, scale inhibition, etc.) depend on nanoscale processes, it is reasonable to expect one outcome of nanotechnology research to be better, nano-engineered water treatment approaches. The most immediate, and possibly greatest, impact of nanotechnology on desalination methods will likely be the development of membranes engineered at the near-molecular level. Aquaporin proteins that channel water across cell membranes with very low energy inputs point to the potential for dramatically improved performance. Aquaporin-laced polymer membranes and aquaporin-mimicking carbon nanotubes and metal oxide membranes developed in the lab support this. A critical limitation to widespread use of nanoengineered desalination membranes will be their scalability to industrial fabrication processes. Subsequent, long-term improvements in nanoengineered membranes may result in self-healing membranes that ideally are (1) more resistant to biofouling, (2) have biocidal properties, and/or (3) selectively target trace contaminants.

The E-Area Low-Level Waste Facility (ELLWF) consists of a number of disposal units described in the Performance Assessment (PA)(WSRC, 2008b) and Composite Analysis (CA)(WSRC, 1997; WSRC, 1999): Low-Activity Waste (LAW) Vault, Intermediate Level (IL) Vault, Trenches (Slit Trenches [STs], Engineered Trenches [ETs], and Component-in-Grout [CIG] Trenches), and Naval Reactor Component Disposal Areas (NRCDAs). This annual review evaluates the adequacy of the approved 2008 ELLWF PA along with the Special Analyses (SAs) approved since the PA was issued. The review also verifies that the Fiscal Year (FY) 2010 low-level waste (LLW) disposal operations were conducted within the bounds of the PA/SA baseline, the Savannah River Site (SRS) CA, and the Department of Energy (DOE) Disposal Authorization Statement (DAS). Important factors considered in this review include waste receipts, results from monitoring and research and development (R&D) programs, and the adequacy of controls derived from the PA/SA baseline. Sections 1.0 and 2.0 of this review are a summary of the adequacy of the PA/SA and CA, respectively. An evaluation of the FY2010 waste receipts and the resultant impact on the ELLWF is summarized in Section 3.1. The results of the monitoring program, R&D program, and other relevant factors are found in Section 3.2, …

Pursuant to the Mercury Export Ban Act of 2008 (P.L. 110-414), DOE was directed to designate a facility or facilities for the long-term management and storage of elemental mercury generated within the United States. Therefore, DOE has analyzed the storage of up to 10,000 metric tons (11,000 tons) of elemental mercury in a facility(ies) constructed and operated in accordance with the Solid Waste Disposal Act, as amended by the Resource Conservation and Recovery Act (74 FR 31723). DOE prepared this Final Mercury Storage EIS in accordance with the National Environmental Policy Act of 1969 (NEPA), as amended (42 U.S.C. 4321 et seq.), the Council on Environmental Quality (CEQ) implementing regulations (40 CFR 1500–1508), and DOE’s NEPA implementing procedures (10 CFR 1021) to evaluate reasonable alternatives for a facility(ies) for the long-term management and storage of elemental mercury. This Final Mercury Storage EIS analyzes the potential environmental, human health, and socioeconomic impacts of elemental mercury storage at seven candidate locations: Grand Junction Disposal Site near Grand Junction, Colorado; Hanford Site near Richland, Washington; Hawthorne Army Depot near Hawthorne, Nevada; Idaho National Laboratory near Idaho Falls, Idaho; Kansas City Plant in Kansas City, Missouri; Savannah River Site near Aiken, South Carolina; and …

This thesis describes a search for a new hadronic resonance in 3.2 fb{sup -1} of data using the Collider Detector at Fermilab. The Fermilab Tevatron accelerator collides beams of protons and antiprotons at a center of mass energy of {radical}s = 1.96 TeV. A unique approach is presented to extract multijet resonances from the large QCD background. Although the search is model independent, a pair produced supersymmetric gluino decaying through R-parity violation into three partons each is used to test our sensitivity to new physics. We measure these partons as jets, and require a minimum of six jets in an event. We make use of the kinematic features and correlations and use an ensemble of jet combinations to distinguish signal from multijet QCD backgrounds. Our background estimates also include all-hadronic t{bar t} decays that have a signature similar to signal. We observe no significant excess in an invariant mass range of 77 GeV/c{sup 2} to 240 GeV/c{sup 2} and place 95% C.L. limits on {sigma}(p{bar p} {yields} {tilde g}{tilde g} {yields} 3jets + 3jets) as a function of gluino invariant mass.

The AGR-1 irradiation experiment ended on November 6, 2009, after 620 effective full power days in the Advanced Test Reactor, achieving a peak burnup of 19.6% FIMA. The test train was shipped to the Materials and Fuels Complex in March 2010 for post-irradiation examination. The first PIE activities included non-destructive examination of the test train, followed by disassembly of the test train and individual capsules and detailed inspection of the capsule contents, including the fuel compacts and the graphite fuel holders. Dimensional measurements of the compacts, graphite holders, and steel capsules shells were performed using a custom vision measurement system (for outer diameters and lengths) and conventional bore gauges (for inner diameters). Gamma spectrometry of the intact test train gave a preliminary look at the condition of the interior components. No evidence of damage to compacts or graphite components was evident from the isotopic and gross gamma scans. Neutron radiography of the intact Capsule 2 showed a high degree of detail of interior components and confirmed the observation that there was no major damage to the capsule. Disassembly of the capsules was initiated using procedures qualified during out-of-cell mockup testing. Difficulties were encountered during capsule disassembly due to irradiation-induced changes …

This report summarizes FY10 accomplishments of the Next Generation Nuclear Plant (NGNP) Engineering Process Heat Applications group in support of hydrogen production technology development. This organization is responsible for systems needed to transfer high temperature heat from a high temperature gas-cooled reactor (HTGR) reactor (being developed by the INL NGNP Project) to electric power generation and to potential industrial applications including the production of hydrogen.

Since understanding and quantifying cosmic ray induced radioactive backgrounds in copper and germanium are important to the MAJORANA DEMONSTRATOR, methods are needed for monitoring the levels of such backgrounds produced in materials being transported and processed for the experiment. This report focuses on work conducted at Pacific Northwest National Laboratory to develop a muon witness system as a one way of monitoring induced activities. The operational goal of this apparatus is to characterize cosmic ray exposure of materials. The cosmic ray flux at the Earth’s surface is composed of several types of particles, including neutrons, muons, gamma rays and protons. These particles induce nuclear reactions, generating isotopes that contribute to the radiological background. Underground, the main mechanism of activation is by muon produced spallation neutrons since the hadron component of cosmic rays is removed at depths greater than a few tens of meters. This is a sub-dominant contributor above ground, but muons become predominant in underground experiments. For low-background experiments cosmogenic production of certain isotopes, such as 68Ge and 60Co, must be accounted for in the background budgets. Muons act as minimum ionizing particles, depositing a fixed amount of energy per unit length in a material, and have a very …

An ideal red phosphor for blue LEDs is one of the biggest challenges for the solid-state lighting industry. The appropriate phosphor material should have good adsorption and emission properties, good thermal and chemical stability, minimal thermal quenching, high quantum yield, and is preferably inexpensive and easy to fabricate. Tantalates possess many of these criteria, and lithium lanthanum tantalate materials warrant thorough investigation. In this study, we investigated red luminescence of two lithium lanthanum tantalates via three mechanisms: (1) Eu-doping, (2) Mn-doping and (3) self-activation of the tantalum polyhedra. Of these three mechanisms, Mn-doping proved to be the most promising. These materials exhibit two very broad adsorption peaks; one in the UV and one in the blue region of the spectrum; both can be exploited in LED applications. Furthermore, Mn-doping can be accomplished in two ways; ion-exchange and direct solid-state synthesis. One of the two lithium lanthanum tantalate phases investigated proved to be a superior host for Mn-luminescence, suggesting the crystal chemistry of the host lattice is important.

Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to …

The Hydrogen Technologies and Systems Center (HTSC) at the U.S. Department of Energy's (DOE) National Renewable Energy Laboratory (NREL) uses a systems engineering and integration approach to hydrogen research and development to help the United States make the transition to a new energy future - a future built on diverse and abundant domestic renewable resources and integrated hydrogen systems. Research focuses on renewable hydrogen production, delivery, and storage; fuel cells and fuel cell manufacturing; technology validation; safety, codes, and standards; analysis; education; and market transformation. Hydrogen can be used in fuel cells to power vehicles and to provide electricity and heat for homes and offices. This flexibility, combined with our increasing demand for energy, opens the door for hydrogen power systems. HTSC collaborates with DOE, other government agencies, industry, communities, universities, national laboratories, and other stakeholders to promote a clean and secure energy future.

This report summarizes the data obtained in a 12-month comparison of three gasoline hybrid electric delivery vehicles with three comparable diesel vehicles. The data show that there was no statistical difference between operating cost per mile of the two groups of vehicles. As expected, tailpipe emissions were considerably lower across all drive cycles for the gHEV than for the diesel vehicle.

The purpose of this report is to provide the basis and the documentation to demonstrate general compliance with the National Emission Standard for Hazardous Air Pollutants (NESHAPS) 40 CFR 61 Subpart H, “National Emission Standards for Emissions of Radionuclides Other Than Radon from Department of Energy Facilities,” (the Standard) for outdoor linear accelerator operations at the Idaho National Laboratory (INL) Stand-off Experiments Range (SOX). The intent of this report is to inform and gain acceptance of this methodology from the governmental bodies regulating the INL.

Power conversion systems for energy storage and other distributed energy resource applications are among the drivers of the important role that power electronics plays in providing reliable electricity. Wide band gap semiconductors such as silicon carbide (SiC) and gallium nitride (GaN) will help increase the performance and efficiency of power electronic equipment while condition monitoring (CM) and prognostics and health management (PHM) will increase the operational availability of the equipment and thereby make it more cost effective. Voltage and/or temperature stress testing were performed on a number of SiC devices in order to accelerate failure modes and to identify measureable shifts in electrical characteristics which may provide early indication of those failures. Those shifts can be interpreted and modeled to provide prognostic signatures for use in CM and/or PHM. Such experiments will also lead to a deeper understanding of basic device physics and the degradation mechanisms behind failure.

The need for a defendable and systematic uncertainty and sensitivity approach that conforms to the Code Scaling, Applicability, and Uncertainty (CSAU) process, and that could be used for a wide variety of software codes, was defined in 2008. The GRS (Gesellschaft für Anlagen und Reaktorsicherheit) company of Germany has developed one type of CSAU approach that is particularly well suited for legacy coupled core analysis codes, and a trial version of their commercial software product SUSA (Software for Uncertainty and Sensitivity Analyses) was acquired on May 12, 2010. This report summarized the results of the initial investigations performed with SUSA, utilizing a typical High Temperature Reactor benchmark (the IAEA CRP-5 PBMR 400MW Exercise 2) and the PEBBED-THERMIX suite of codes. The following steps were performed as part of the uncertainty and sensitivity analysis: 1. Eight PEBBED-THERMIX model input parameters were selected for inclusion in the uncertainty study: the total reactor power, inlet gas temperature, decay heat, and the specific heat capability and thermal conductivity of the fuel, pebble bed and reflector graphite. 2. The input parameters variations and probability density functions were specified, and a total of 800 PEBBED-THERMIX model calculations were performed, divided into 4 sets of 100 and …

During FY 2011 the CP & DB Program will report Highlights on a monthly basis, but will no longer produce Quarterly Progress Reports. Technical details that were previously included in the quarterly reports will be included in the appropriate Milestone Reports that are submitted to FCRD Program Management. These reports will also be uploaded to the Deep Burn website. The Monthly Highlights report for November 2010, ORNL/TM-2010/323, was distributed to program participants on December 9, 2010. The final Quarterly for FY 2010, Deep Burn Program Quarterly Report for July - September 2010, ORNL/TM-2010/301, was announced to program participants and posted to the website on December 28, 2010. This report discusses the following: (1) Thermochemical Data and Model Development - (a) Thermochemical Modeling, (b) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Pebble Bed Design (INL), (c) Radiation Damage and Properties; (2) TRISO (tri-structural isotropic) Development - (a) TRU (transuranic elements) Kernel Development, (b) Coating Development; (3) LWR Fully Ceramic Fuel - (a) FCM Fabrication Development, (b) FCM Irradiation Testing (ORNL); (4) Fuel Performance and Analytical Analysis - Fuel Performance Modeling (ORNL).

This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the …

The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility, the highest ranked alternative, will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

Sandia National Laboratories and General Motors Global Energy Systems team conducted a joint biofuels systems analysis project from March to November 2008. The purpose of this study was to assess the feasibility, implications, limitations, and enablers of large-scale production of biofuels. 90 billion gallons of ethanol (the energy equivalent of approximately 60 billion gallons of gasoline) per year by 2030 was chosen as the book-end target to understand an aggressive deployment. Since previous studies have addressed the potential of biomass but not the supply chain rollout needed to achieve large production targets, the focus of this study was on a comprehensive systems understanding the evolution of the full supply chain and key interdependencies over time. The supply chain components examined in this study included agricultural land use changes, production of biomass feedstocks, storage and transportation of these feedstocks, construction of conversion plants, conversion of feedstocks to ethanol at these plants, transportation of ethanol and blending with gasoline, and distribution to retail outlets. To support this analysis, we developed a 'Seed to Station' system dynamics model (Biofuels Deployment Model - BDM) to explore the feasibility of meeting specified ethanol production targets. The focus of this report is water and its linkage …

Dust is generated in fusion reactors from plasma erosion of plasma facing components within the reactor’s vacuum vessel (VV) during reactor operation. This dust collects in cooler regions on interior surfaces of the VV. Because this dust can be radioactive, toxic, and/or chemically reactive, it poses a safety concern, especially if mobilized by the process of resuspension during an accident and then transported as an aerosol though out the reactor confinement building, and possibly released to the environment. A computer code used at the Idaho National Laboratory (INL) to model aerosol transport for safety consequence analysis is the MELCOR code. A primary reason for selecting MELCOR for this application is its aerosol transport capabilities. The INL Fusion Safety Program (FSP) organization has made fusion specific modifications to MELCOR. Recent modifications include the implementation of aerosol resuspension models in MELCOR 1.8.5 for Fusion. This paper presents the resuspension models adopted and the initial benchmarking of these models.

The purpose of this LDRD was to generate data that could be used to populate and thereby reduce the uncertainty in global carbon cycle models. These efforts were focused on developing a system for determining the dissolution rate of biogenic calcite under oceanic pressure and temperature conditions and on carrying out a digital transcriptomic analysis of gene expression in response to changes in pCO2, and the consequent acidification of the growth medium.

Roman glasses that have been in the sea or underground for about 1800 years can serve as the independent “experiment” that is needed for validation of codes and models that are used in performance assessment. Two sets of Roman-era glasses have been obtained for this purpose. One set comes from the sunken vessel the Iulia Felix; the second from recently excavated glasses from a Roman villa in Aquileia, Italy. The specimens contain glass artifacts and attached sediment or soil. In the case of the Iulia Felix glasses quite a lot of analytical work has been completed at the University of Padova, but from an archaeological perspective. The glasses from Aquileia have not been so carefully analyzed, but they are similar to other Roman glasses. Both glass and sediment or soil need to be analyzed and are the subject of this analytical plan. The glasses need to be analyzed with the goal of validating the model used to describe glass dissolution. The sediment and soil need to be analyzed to determine the profile of elements released from the glass. This latter need represents a significant analytical challenge because of the trace quantities that need to be analyzed. Both pieces of information …

ABSTRACT The HTR Module [1] is a graphite-moderated, helium cooled pebble bed High Temperature Reactor (HTR) design that has been extensively used as a reference template for the former South African and current Chinese HTR [2] programs. This design utilized spherical fuel elements packed into a dynamic pebble bed, consisting of TRISO coated uranium oxide (UO2) fuel kernels with a U-235 enrichment of 7.8% and a Heavy Metal loading of 7 grams per pebble. The main objective of this study is to compare several important reactor physics and core design parameters for the HTR Module and an identical design utilizing UCO fuel kernels. Fuel kernels of this type are currently being tested in the Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR) as part of the larger Next Generation Nuclear Plant (NGNP) project. The PEBBED-THERMIX [3] code, which was developed specifically for the analysis of pebble bed HTRs, was used to compare the coupled neutronic and thermal fluid performance of the two designs.

The Intergovernmental Panel on Climate Change (IPCC) Fourth Assessment Report (AR4) made extensive use of coordinated simulations by 18 international modeling groups using a variety of coupled general circulation models (GCMs) with different numerics, algorithms, resolutions, physics models, and parameterizations. These simulations span the 20th century and provide forecasts for various carbon emissions scenarios in the 21st century. All the output from this panoply of models is made available to researchers on an archive maintained by the Program for Climate Model Diagnosis and Intercomparison (PCMDI) at LLNL. I have downloaded this data and completed the first steps toward a statistical analysis of these ensembles for the US Southwest. This constitutes the final report for a late start LDRD project. Complete analysis will be the subject of a forthcoming report.

New and enhanced nuclear fuels are a key enabler for new and improved reactor technologies. For example, the goals of the next generation nuclear plant (NGNP) will not be met without irradiations successfully demonstrating the safety and reliability of new fuels. Likewise, fuel reliability has become paramount in ensuring the competitiveness of nuclear power plants. Recently, the Office of Nuclear Energy in the Department of Energy (DOE-NE) launched a new direction in fuel research and development that emphasizes an approach relying on first principle models to develop optimized fuel designs that offer significant improvements over current fuels. To facilitate this approach, high fidelity, real-time, data are essential for characterizing the performance of new fuels during irradiation testing. A three-year strategic research program is proposed for developing the required test vehicles with sensors of unprecedented accuracy and resolution for obtaining the data needed to characterize three-dimensional changes in fuel microstructure during irradiation testing. When implemented, this strategy will yield test capsule designs that are instrumented with new sensor technologies for the Advanced Test Reactor (ATR) and other irradiation locations for the Fuel Cycle Research and Development (FC R&D) program. Prior laboratory testing, and as needed, irradiation testing, of these sensors will …