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It has recently been suggested that in certain special nonsupersymmetric type II string compactifications, at least the first two perturbative contributions to the cosmological constant {Lambda} vanish. Support for perturbative vanishing beyond 1-loop (as well as evidence for the absence of some nonperturbative contributions) has come from duality arguments. There was also a direct 2-loop computation which was incomplete; in this note we explain the deficiency of the previous 2-loop calculation and discuss the complete 2-loop computation in two different models. The corrected analysis yields a vanishing 2-loop contribution to {Lambda} in these models.

Electrometallurgical treatment is a compact, inexpensive method that is being developed at Argonne National Laboratory to deal with spent nuclear fuel, primarily metallic and oxide fuels. In this method, metallic nuclear fuel constituents are electrorefined in a molten salt to separate uranium from the rest of the spent fuel. Oxide and other fuels are subjected to appropriate head end steps to convert them to metallic form prior to electrorefining. The treatment process generates two kinds of high-level waste--a metallic and a ceramic waste. Isolation of these wastes has been developed as an integral part of the process. The wastes arise directly from the electrorefiner, and waste streams do not contain large quantities of solvent or other process fluids. Consequently, waste volumes are small and waste isolation processes can be compact and rapid. This paper briefly summarizes waste isolation processes then describes development and characterization of the two waste forms in more detail.

The DART code is based upon a thermomechanical model that can predict swelling, recrystallization, fuel-meat interdiffusion and other issues related with MTR dispersed FE behavior under irradiation. As a part of a common effort to develop an optimized version of DART, a comparison between DART predictions and CNEA miniplates irradiation experimental data was made. The irradiation took place during 1981-82 for U3O8 miniplates and 1985-86 for U{sub 3}Si{sub x} at Oak Ridge Research Reactor (ORR). The microphotographs were studied by means of IMAWIN 3.0 Image Analysis Code and different fission gas bubbles distributions were obtained. Also it was possible to find and identify different morphologic zones. In both kinds of fuels, different phases were recognized, like particle peripheral zones with evidence of Al-U reaction, internal recrystallized zones and bubbles. A very good agreement between code prediction and irradiation results was found. The few discrepancies are due to local, fabrication and irradiation uncertainties, as the presence of U{sub 3}Si phase in U{sub 3}Si{sub 2} particles and effective burnup.

The structure of a completely amorphous zircon was determined by time-of-flight neutron diffraction at Argonne's Intense Pulsed Neutron Source (IPNS). The sample of metamict zircon (ZrSiO{sub 4}),initially doped to 8.85 weight percent {sup 238}Pi, had been completely amorphized by alpha-recoil damage since its synthesis in 1981 at the Pacific Northwest National Laboratory (PNNL). The measured diffraction structure factor, S(Q), indicated a completely amorphous sample, with no signs of residual zircon microcrystallinity. The pair distribution function obtained indicated that the structure was that of an oxide glass, retaining the Si-0, Zr-0, and O-O bond lengths of crystalline zircon.

Various alternative fuels and improved engine and vehicle systems have been proposed in order to reduce emissions and energy use associated with heavy vehicles (predominantly trucks). For example, oil companies have proposed improved methods for converting natural gas to zero-aromatics, zero-sulfur diesel fuel via the Fischer-Tropsch process. Major heavy-duty diesel engine companies are working on ways to simultaneously reduce particulate-matter and NOX emissions. The trend in heavy vehicles is toward use of lightweight materials, tires with lower rolling resistance, and treatments to reduce aerodynamic drag. In this paper, we compare the Mecycle energy use and emissions from trucks using selected alternatives, such as Fisher-Tropsch diesel fuel and advanced fuel-efficient engines. We consider heavy-duty, Class 8 tractor-semitrailer combinations for this analysis. The total life cycle includes production and recycling of the vehicle itself, extraction, processing, and transportation of the fuel itself, and vehicle operation and maintenance. Energy use is considered in toto, as well as those portions that are imported, domestic, and renewable. Emissions of interest include greenhouse gases and criteria pollutants. Angonne's Greenhouse Gases, Regulated Emissions, and Energy Use in Transportation (GREET) model is used to generate per-vehicle fuel cycle impacts. Energy use and emissions for materials manufacturing and vehicle …

At an international conference on global warming, held in Kyoto, Japan, in December 1997, the United States committed to reduce its greenhouse gas (GHG) emissions by 7% over its 1990 level by the year 2012. To help achieve that goal, transportation GHG emissions need to be reduced. Using Argonne's fuel-cycle model, I estimated GHG emissions reduction potentials of various near- and long-term transportation technologies. The estimated per-mile GHG emissions results show that alternative transportation fuels and advanced vehicle technologies can help significantly reduce transportation GHG emissions. Of the near-term technologies evaluated in this study, electric vehicles; hybrid electric vehicles; compression-ignition, direct-injection vehicles; and E85 flexible fuel vehicles can reduce fuel-cycle GHG emissions by more than 25%, on the fuel-cycle basis. Electric vehicles powered by electricity generated primarily from nuclear and renewable sources can reduce GHG emissions by 80%. Other alternative fuels, such as compressed natural gas and liquefied petroleum gas, offer limited, but positive, GHG emission reduction benefits. Among the long-term technologies evaluated in this study, conventional spark ignition and compression ignition engines powered by alternative fuels and gasoline- and diesel-powered advanced vehicles can reduce GHG emissions by 10% to 30%. Ethanol dedicated vehicles, electric vehicles, hybrid electric vehicles, and …

A series of studies is presented in which an electrical-impedance tomography (EXT) system is validated for two-phase flow measurements. The EIT system, developed at Sandia National Laboratories, is described along with the computer algorithm used for reconstructing phase volume fraction profiles. The algorithm is first tested using numerical data and experimental phantom measurements, with good results. The EIT system is then applied to solid-liquid and gas-liquid flows, and results are compared to an established gamma-densitometry tomography (GDT) system. In the solid-liquid flows, the average solid volume fractions measured by EIT are in good agreement with nominal values; in the gas-liquid flows, average gas volume fractions and radial gas volume fraction profiles from GDT and EIT are also in good agreement.

The results from the RELAP5/MOD3 and PARET/ANL codes are compared with the SPERT-IV series of experimental reactivity insertion transients. The PARET/ANL code provides conservative estimates of SPERT-IV experimental data for the midrange transients and for the more severe transients. The PARET results are similar to the results obtained earlier for the SPERT-I D-12/25 series of experiments. The RELAP5/MOD3 code (including the developmental version 3.2.1.2) gives results comparable to PARET for some midrange transients, but seriously diverges from the experimental data when significant boiling is present. Based on the results of this study, the use of the RELAP5 code for research reactor applications should be limited to transients that do not generate substantial boiling and voids. We hope to be able to resolve these differences in further work with the NRC staff and its contractors. The RELAP5 code would be a more useful tool for the analyses research reactor transients with the addition of suitable correlations for low pressures and plate type geometry.

The Self-Describing Data Sets (SDDS) toolkit comprises about 70 generally-applicable programs sharing a common data protocol. At the Advanced Photon Source (APS), SDDS performs the vast majority of operational data collection and processing, most data display functions, and many control functions. In addition, a number of accelerator simulation codes use SDDS for all post-processing and data display. This has three principle advantages: first, simulation codes need not provide customized post-processing tools, thus simplifying development and maintenance. Second, users can enhance code capabilities without changing the code itself, by adding SDDS-based pre- and post-processing. Third, multiple codes can be used together more easily, by employing SDDS for data transfer and adaptation. Given its broad applicability, the SDDS file protocol is surprisingly simple, making it quite easy for simulations to generate SDDS-compliant data. This paper discusses the philosophy behind SDDS, contrasting it with some recent trends, and outlines the capabilities of the toolkit. The paper also gives examples of using SDDS for accelerator simulation.

The authors have performed the reaction {sup 248}Cm({sup 4}He, {sup 3}He) using 98.5-MeV alpha particles from the IUCF cyclotron to populate high-j states in {sup 249}Cm. A tentative assignment of the K{sub 17/2} component of the 1/2{sup +}[880] Nilsson state has been made.

The dielectric wakefield transformer (DWT) is one route to practical high energy wakefield-based accelerators. Progress has been made in a number of areas relevant to the demonstration of this device. In this article we describe recent bench measurements and beam experiments using 7.8 and 15.6 GHz structures and discuss some remaining technical challenges in the development of the DWT.

A neutronic feasibility study for converting the High Flux Beam Reactor at Brookhaven National Laboratory from HEU to LEU fuel was performed at Argonne National Laboratory. The purpose of this study is to determine what LEU fuel density would be needed to provide fuel lifetime and neutron flux performance similar to the current HEU fuel. The results indicate that it is not possible to convert the HFBR to LEU fuel with the current reactor core configuration. To use LEU fuel, either the core needs to be reconfigured to increase the neutron thermalization or a new LEU reactor design needs to be considered. This paper presents results of reactor calculations for a reference 28-assembly HEU-fuel core configuration and for an alternative 18-assembly LEU-fuel core configuration with increased neutron thermalization. Neutronic studies show that similar in-core and ex-core neutron fluxes, and fuel cycle length can be achieved using high-density LEU fuel with about 6.1 gU/cm{sup 3} in an altered reactor core configuration. However, hydraulic and safety analyses of the altered HFBR core configuration needs to be performed in order to establish the feasibility of this concept.

Argonne National Laboratory (ANL-E) has recently completed the decontamination and decommissioning (D and D) of the JANUS Reactor Facility located in Building 202. The 200 KW reactor operated from August 1963 to March 1992. The facility was used to study the effects of both high and low doses of fission neutrons in animals. There were two exposure rooms on opposite sides of the reactor and the reactor was therefore named after the two-faced Roman god. The High Dose Room was capable of specimen exposure at a dose rate of 3,600 rads per hour. During calendar year 1996 a detailed characterization of the facility was performed by ANL-E Health Physics personnel. ANL-E Analytical Services performed the required sample analysis. An Auditable Safety Analysis and an Environmental Assessment were completed. D and D plans, procedures and procurement documents were prepared and approved. A D and D subcontractor was selected and a firm, fixed price contract awarded for the field work and final survey effort. The D and D subcontractor was mobilized to ANL-E in January 1997. Electrical isolation of all reactor equipment and control panels was accomplished and the equipment removed. A total of 207,230 pounds (94,082 Kg) of lead shielding was …

The Field Lysimeter Investigations: Low-level Waste Data Base Development Program is obtaining information on the performance of radioactive waste forms. These experiments were recently shut down and have been examined in accordance with a detailed waste form and soil sampling plan. Ion-exchange resins from a commercial nuclear power station were solidified into waste forms using portland cement and vinyl ester-styrene. These waste forms were tested to (a) obtain information on performance of waste forms in typical disposal environments, (b) compare field results with bench leach studies, (c) develop a low-level waste data base for use in performance assessment source term calculations, and (d) apply the DUST computer code to compare predicted cumulative release to actual field data. The program, funded by the Nuclear Regulatory Commission (NRC), includes observed radionuclide releases from waste forms at two test sites over 10 years of successful operation. The lysimeter is a useful experimental device that is providing source term data and can be used to determine waste form stability. Lysimeters are ideal systems for obtaining actual field test data because, when properly designed and operated, they can be used to isolate soil and waste systems under actual environmental conditions. Such conditions cannot be duplicated …

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This paper presents the status of accelerator facilities in North America that are involved in research using radioactive nuclear beams (RNB), including existing and operating facilities, ones currently under construction or undergoing major upgrades, and ones being planned or proposed for the future. Existing RNB facilities are located at TRIUMF (TISOL) in Vancouver, B.C., the Holifield Radioactive Ion Beam Facility (HRIBF) at Oak Ridge National Laboratory, the Argonne Tandem Linear Accelerator System (ATLAS) at Argonne National Laboratory, the National Superconducting Cyclotron Laboratory (NSCL) at Michigan State University, the Nuclear Structure Laboratory at The University of Notre Dame, the 88 Cyclotron at Lawrence Berkeley National Laboratory, and the Cyclotron Institute at Texas A and M University. Currently, there are two major RNB facility upgrades in progress in North America, one at TRIUMF, the ISAC project, and one at NSCL the Intensity Upgrade project. For the future the U. S. Nuclear Science Advisory Committee has given high priority for an advanced RNB facility of the ISOL type. Concepts for such a facility, currently being developed at Argonne National Laboratory and Oak Ridge National Laboratory, are presented. Plans are also being developed in Canada at TRIUMF for a major upgrade of the ISAC …

One of the major projects being considered for future construction at Fermilab is a Very Large Hadron Collider (VLHC). This project will require a new High Energy Booster ring, 33 km in circumference, to provide 3 TeV proton beams for injection into the VLHC. Here we discuss the possibility to build an ep collider in the VLHC Booster tunnel. A LEP-like e{sup {+-}} beam (83 GeV) would collide with one of the proton beams (3 TeV) to create a {radical}s {approx} 1 TeV ep collider. Design groups at Argonne and Fermilab have established that it is feasible to build electron injection at the existing infrastructure, and provide a machine with luminosity exceeding 10{sup 32} cm{sup {minus}2} s{sup {minus}1}.

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. Even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification of available models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The Working Group agreed that each side would irradiate, anneal, reirradiate (if feasible ), and test two materials of the other. Charpy V-notch (CVN) and tensile specimens were included. Oak Ridge National Laboratory (ORNL) conducted such a program (irradiation and annealing, including static fracture toughness) with two weld metals representative of VVER-440 and VVER-1000 RPVs, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation, annealing, reirradiation, …

Minimization of component activation is highly desirable at accelerator-based positron sources. Electrons in the 8- to 14-MeV energy range impinging on a target produce photons energetic enough to create electron-positron pairs; however, few of the photons are energetic enough to produce photoneutrons. Slow positron production by low-energy electrons impinging on a multilayer tungsten target with and without electromagnetic extraction between the layers was studied by simulation. The neutron background from 14-MeV electrons is expected to be significantly lower than that encountered with higher-energy electron beams. Numerical results are presented and some ideas for a low-activation slow-positron source are discussed.

Argonne National Laboratory is developing an electrometallurgical treatment for spent fuel from the experimental breeder reactor II. A product of this treatment process is a metal waste form that incorporates the stainless steel cladding hulls, zirconium from the fuel and the fission products that are noble to the process, i.e., Tc, Ru, Pd, Rh, Ag. The nominal composition of this waste form is stainless steel/15 wt% zirconium 1-4 wt% noble metal fission products. The behavior of technetium is of particular importance from a disposal point of view for this waste form due to its long half life, 2. 14E5 years, and its mobility in groundwater. To address these concerns a limited number of spiked metal waste forms were produced containing Tc. These surrogate waste forms were then studied using scanning electron microscopy and selected leaching tests.

Article on a computational investigations of iodine oxides.

A multigrid strategy is developed for accelerating the steady state computations of waves propagating with curvature dependent speeds. This will allow the rapid computation of a burn table. In a high explosive material, the creation of a burn table will allow the elimination of solving chemical reaction ODEs and feed in source terms to the reactive flow equations for solution of the system of ignition of the high explosive material. Standard iterative methods show a quick reduction of the residual followed by a slow final convergence to the solution at high iterations. Such systems are excellent choices for the use of multigrid methods to speed up convergence, even on a nonlinear system such as this. Numerical steady-state solutions to the eikonal equation on a rectangular grid are conducted. Results are presented for a square grid in 2D and a cubic grid in 3D using a Runge-Kutta time iteration for the smoothing operator until steady-state is reached.

The results of molecular dynamics (MD) displacement cascade simulations in bcc iron have been used to obtain effective cross sections for two measures of primary damage production: (1) the number of surviving point defects expressed as a fraction of the displacements calculated using the standard secondary displacement model of Norgett, Robinson, and Torrens (NRT), and (2) the fraction of the surviving interstitials contained in clusters that formed during the cascade event. Primary knockon atom spectra for iron obtained from the SPECTER code have been used to weight these MD-based damage production cross sections in order to obtain spectrally-averaged values for several locations in commercial fission reactors and materials test reactors. An evaluation of these results indicates that neutron energy spectrum differences between the various enviromnents do not lead to significant differences between the average primary damage formation parameters. In particular, the defect production cross sections obtained for PWR and BWR neutron spectra were not significantly different. The variation of the defect production cross sections as a function of depth into the reactor pressure vessel wall is used as a sample application of the cross sections. A slight difference between the attenuation behavior of the PWR and BWR was noted; this …