Publications

The corrosion behavior of a rapidly solidified Al-Fe-Gd alloy glass was studied in aqueous chloride environments of varying pH using potentiodynamic polarization. The corrosion behavior of a rapidly solidified pure Al glass, crystalline Al-Fe-Gd alloy and crystalline Al were measured for comparison. Due to lattice disorder, the dissolution rate of the glasses in the passive region was greater than that of their crystalline counterparts. The breakaway potentials measured for the glasses were more positive than those of the crystalline metals because film breakdown initiation sites like second phase particles and internal boundaries were not present in the glass. Alloy glass specimens exposed to alkaline solutions exhibited lower passive current densities and higher breakaway potentials than expected. This appeared to be a result of an enrichment of oxidized Fe at the specimen surface. This Fe-rich protective film also appeared to form in pits which developed on crystallized specimens tested in neutral chloride solutions.

There are tremendous resources of natural gas in tight fissured rocks, but these formations require special care for hydraulic fracturing to be successful. Serious problems include leakoff, damage and complex fracturing. Leakoff may be constant, pressure-sensitive, or accelerating. Leakoff becomes most severe when fissures begin to dilate and accept large volumes of fracture fluid, which may rapidly dehydrate a sandladen slurry. Determining values of pressure-sensitive and accelerated leakoff coefficients is difficult, and generally requires both a pressure-decline analysis (after a minifrac) and an analysis of the injection pressure. Fine-mesh sand, often used in fissured reservoirs, will help control leakoff. Damage to the natural fractures, due to leakoff of the fluid and gels and to mechanical types of damage, must be avoided, since the fissures are the production mechanism. It is particularly important to minimize the amount of liquid and gels, since the fissures are narrow and easily blocked. These concepts are also applicable to oil reservoirs.

This document contains the steady-state and loss-of-pumping accident analysis of the representative design for the Savannah River heavy water new production reactor. A description of the reactor system and computer input model, the results of the steady-state analysis, and the results of four loss-of-pumping accident calculations are presented. 5 refs., 37 figs., 4 tabs.

This report lists the measurement capabilities of the Department of Energy Contractors' Standard Laboratories within the Nuclear Weapon Complex. It is intended to foster cross-utilization of measurements between laboratories and provides a guide for survey and audit activities. Although this report was prepared by Sandia Primary Standards Laboratory (PSL), the PSL was intentionally omitted. Capabilities of the PSL are documented in SAND88-3402.UC-700.

Early attempts at estimation of stress wave damage due to blasting by use of finite element calculations met with limited success due to numerical instabilities that prevented calculations from being carried to late times. An improved damage model allows finite element calculations which remain stable at late times. Reasonable agreement between crater profiles calculated with this model using the PRONTO finite element program and excavated crater profiles from blasting experiments in granite demonstrate a successful application of this model. Detailed instructions for use of this new damage model with the PRONTO finite element programs are included. 18 refs., 16 figs.

The DEBRIS module was developed to deal with the analysis of core melt processes in Light Water Reactors (LWR's). It was designed to address the important processes associated with the late phase'' of a core meltdown. This phase encompasses the period following the loss of intact rod geometry and ending with vessel head failure. It is characterized by the melting and relocation of ceramic rich materials through a rubblized medium composed primarily of fuel pellet and oxidized cladding fragments. Of particular interest are the dynamics of the melting process, the relocation of the components, the formation of crusts, retention of molten materials by the crust, and remelting of crusts. The DEBRIS module treats these processes in a two-dimensional (r,z) geometry solving the continuity, momentum, and energy equations to describe the dynamics of meltdown. The DEBRIS models are described together with some of the analyses to which the module has been applied. In particular, a description is given of the DEBRIS module analysis of the MP-1 experiment. The DEBRIS module appears to have significant potential for the analysis of late phase'' meltdown processes and can be effectively used both in a stand-alone mode or in conjunction with the severe accident analysis codes (MELCOR,SCDAP). In addition, the module may prove effective for treatment of the early phase processes as well. 19 refs., 16 figs.

Position location is a fundamental requirement in autonomous mobile robots which record and subsequently follow x,y paths. The Dept. of Energy, Office of Safeguards and Security, Robotic Security Vehicle (RSV) program involves the development of an autonomous mobile robot for patrolling a structured exterior environment. A straight-forward method for autonomous path-following has been adopted and requires digitizing'' the desired road network by storing x,y coordinates every 2m along the roads. The position location system used to define the locations consists of a radio beacon system which triangulates position off two known transponders, and dead reckoning with compass and odometer. This paper addresses the problem of combining these two measurements to arrive at a best estimate of position. Two algorithms are proposed: the optimal'' algorithm treats the measurements as random variables and minimizes the estimate variance, while the average error'' algorithm considers the bias in dead reckoning and attempts to guarantee an average error. Data collected on the algorithms indicate that both work well in practice. 2 refs., 7 figs.

The MC4039 Unique Signal Override Plug (USOP) provides the unique signal for the B90 when fielded on aircraft that are not equipped with unique signal capability. Since the USOP is field installed, the concern is that it might be susceptible to electromagnetic radiation prior to installation on the weapon. This report documents a characterization of the USOP, evaluates various techniques for attaching electromagnetic shields, and evaluates the susceptibility of a fully assembled passive-USOP. Tests conducted evaluated the electromagnetic susceptibility of the passive, unconnected USOP. During normal operation the USOP is powered directly from the weapon. During the course of this test program two prototypes were developed. The prototype 1 USOP internal circuitry contains one SA3727 chip, five diodes, three resistors, and two capacitors; these are mounted on a circular circuit board and contained inside a metal back shell cover, which serves as an electromagnetic shield. The prototype 2 design incorporated four changes. The manufacturer of the SA3727 chip was changed from Lasarray to LSI Logic, the circuit board ground was tied to the case ground through a straight wire, Cl was changed from 1 microfarad to 0.1 microfarads. and the circuit board was changed, as required. 2 refs., 17 figs., 3 tabs. (JF)

Using Sequoyah as a representative plant, calculations have been performed with a developmental version of the CONTAIN computer code to assess the effectiveness of various possible improvements to ice condenser containments in mitigating severe accident scenarios involving direct containment heating (DCH) and/or hydrogen combustion. Mitigation strategies considered included backup power for igniters and/or air return fans, augmented igniter systems, containment venting, containment inerting, subatmospheric containment operation, reduced ice condenser bypass, and primary system depressurization. Various combinations of these improvements were also considered. Only inerting the containment or primary system depressurization combined with backup power supplies for the igniter systems resulted in large decreases in the peak pressures calculated to result from DCH events. Potential hydrogen detonation threats were also assessed; providing backup power for both the igniter systems and the air return fans would significantly reduce the potential for detonations but might not totally eliminate it. Sensitivity studies using the NUREG-1150 PRA methodology indicated that primary system depressurization combined with backup power for both igniters and fans could reduce the contribution to the mean risk potential of the class of events considered by about a factor of three. 7 refs., 6 figs., 6 tabs.

Finding new and improved means of cooling small electronic packages are of great importance to today's electronic packaging engineer. Thermal absorption through the use of a material which changes phase is an attractive alternative. Taking advantage of the heat capacity of a material's latent heat of fusion is shown to absorb heat away from the electronics, thus decreasing the overall temperature rise of the system. The energy equation is formulated in terms of enthalpy and discretized using a finite-difference method. A FORTRAN program to solve the discretized equations is presented which can be used to analyze heat conduction in a rectangular region undergoing an isothermal phase change. An analysis of heat transfer through a miniature radar electronic module cooled by a phase-change reservoir is presented, illustrating the method's advantages over conventional heat sinks. 41 refs., 11 figs., 2 tabs.

This report gives an annotated bibliography of reports published in 1989 by the Nuclear Energy Technology Directorate. A listing is also given of reports published by the staff in the nuclear energy field since 1972.

Remote automated cask handling has the potential to reduce both the occupational exposure and the time required to process a nuclear waste transport cask at a handling facility. The ongoing Advanced Handling Technologies Project (AHTP) at Sandia National Laboratories is described. AHTP was initiated to explore the use of advanced robotic systems to perform cask handling operations at handling facilities for radioactive waste, and to provide guidance to cask designers regarding the impact of robotic handling on cask design. The proof-of-concept robotic systems developed in AHTP are intended to extrapolate from currently available commercial systems to the systems that will be available by the time that a repository would be open for operation. The project investigates those cask handling operations that would be performed at a nuclear waste repository facility during cask receiving and handling. The ongoing AHTP indicates that design guidance, rather than design specification, is appropriate, since the requirements for robotic handling do not place severe restrictions on cask design but rather focus on attention to detail and design for limited dexterity. The cask system design features that facilitate robotic handling operations are discussed, and results obtained from AHTP design and operation experience are summarized. The application of these design considerations is illustrated by discussion of the robot systems and their operation on cask feature mock-ups used in the AHTP project. 11 refs., 11 figs.

This report describes the clutter model developed at Sandia National Laboratories for the SRIM code version 2.2s. The SNL clutter model is a fully polarimetric model that includes both coherent and incoherent scattering effects. The input parameters to the SNL clutter model are chosen so that an acceptable match is obtained between the model predicted data and the appropriate experimental data. These input parameters are then used in the SRIM code to simulated the desired clutter type. 12 refs., 13 figs., 2 tabs.

The DF-4 in-pile fuel damage experiment investigated the behavior of boiling water reactor (BWR) fuel canisters and control blades in the high temperature environment of an unrecovered reactor accident. This experiment, which was carried out in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories, was performed under the USNRC's internationally sponsored severe fuel damage (SFD) program. The DF-4 test is described herein and results from the experiment are presented. Important findings from the DF-4 test include the low temperature melting of the stainless steel control blade caused by reaction with the B{sub 4}C, and the subsequent low temperature attack of the Zr-4 channel box by the relocating molten blade components. Hydrogen generation was found to continue throughout the experiment, diminishing slightly following the relocation of molten oxidizing zircaloy to the lower extreme of the test bundle. A large blockage which was formed from this material continued to oxidize while steam was being fed into the the test bundle. The results of this test have provided information on the initial stages of core melt progression in BWR geometry involving the heatup and cladding oxidation stages of a severe accident and terminating at the point of melting and relocation of the metallic core components. The information is useful in modeling melt progression in BWR core geometry, and provides engineering insight into the key phenomena controlling these processes. 12 refs., 12 figs.

A description is presented of MIDAS, the Mobile Intrusion Detection and Assessment System. MIDAS is a security system that can be quickly deployed to provide wide area coverage for a mobile asset. MIDAS uses two passive infrared imaging sensors, one for intruder detection and one for assessment. Detected targets are tracked while assessment cameras are directed to view the intruder location for operator observation and assessment. The dual sensor design allows simultaneous detection, assessment, and tracking. Control and status information is provided to an operator using a color graphics terminal, touch panel driven menus, and a joystick for control of the assessment sensor pan and tilt.

The Hydrologic Code Intercomparison Project (HYDROCOIN) was formed to evaluate hydrogeologic models and computer codes and their use in performance assessment for high-level radioactive-waste repositories. This report describes the results of a study for HYDROCOIN of model sensitivity for isothermal, unsaturated flow through layered, fractured tuffs. We investigated both the types of flow behavior that dominate the performance measures and the conditions and model parameters that control flow behavior. We also examined the effect of different conceptual models and modeling approaches on our understanding of system behavior. The analyses included single- and multiple-parameter variations about base cases in one-dimensional steady and transient flow and in two-dimensional steady flow. The flow behavior is complex even for the highly simplified and constrained system modeled here. The response of the performance measures is both nonlinear and nonmonotonic. System behavior is dominated by abrupt transitions from matrix to fracture flow and by lateral diversion of flow. The observed behaviors are strongly influenced by the imposed boundary conditions and model constraints. Applied flux plays a critical role in determining the flow type but interacts strongly with the composite-conductivity curves of individual hydrologic units and with the stratigraphy. One-dimensional modeling yields conservative estimates of distributions of groundwater travel time only under very limited conditions. This study demonstrates that it is wrong to equate the shortest possible water-travel path with the fastest path from the repository to the water table. 20 refs., 234 figs., 10 tabs.

The US Department of Energy is responsible for the design, construction, operation, and decommission of a site for the deep geologic disposal of high-level radioactive waste (HLW). This involves site characterization and the use of performance assessment to demonstrate compliance with regulations for HLW disposal from the US Environmental Protection Agency (EPA) and the US Nuclear Regulatory Commission. The EPA standard states that a performance assessment should consider the associated uncertainties involved in estimating cumulative release of radionuclides to the accessible environment. To date, the majority of the efforts in uncertainty analysis have been directed toward data and parameter uncertainty, whereas little effort has been made to treat model uncertainty. Model uncertainty includes conceptual model uncertainty, mathematical model uncertainty, and any uncertainties derived from implementing the mathematical model in a computer code. Currently there is no systematic approach that is designed to address the uncertainty in conceptual models. The purpose of this investigation is to take a first step at addressing conceptual model uncertainty. This will be accomplished by assessing the relative impact of alternative conceptual models on the integrated release of radionuclides to the accessible environment for an HLW repository site located in unsaturated, fractured tuff. 4 refs., 2 figs.

An intermediate step in the development of inertial confinement fusion (ICF) for power production will be the development and testing of a high-gain facility. One concept being considered for this facility is the Laboratory Microfusion Facility (LMF). Other projected applications of the LMF include high-energy-density physics experiments and nuclear effects testing. At the Air Force Institute of Technology (AFIT), Lawrence Livermore National Laboratory (LLNL), and Sandia National Laboratories Albuquerque (SNLA), we have been studying the use of the LMF for nuclear effects experiments. Because of the amount of energy that will be released in a high-gain ICF test and the size of the LMF, test objects could be any size from very small electronic components to large systems; and nuclear effects in the LMF may include thermal radiation, x-rays, electromagnetic pulse, gamma rays, neutrons, or others. At AFIT, LLNL, and SNLA we have been investigating a test to expose systems to high-energy x-ray pulses, and have completed predictions of x-ray fluence, dose, etc. in various conceptual LMF reactors. However, comparison of our results is more meaningful if our prediction methods produce the same results for the same design. To establish a basis for comparison, we set up a simple benchmark problem and we each computed x-ray and neutron transport. The model and codes are described and the results are compared and discussed.

Partial least squares multivariate calibration methods were applied to the infrared spectra of a new set of borophosphosilicate glass (BPSG) thin films on silicon wafers. The calibration samples were prepared by a low pressure chemical vapor deposition (LPCVD) process. The statistically designed calibration set included data from nearly 400 coated Si wafers. Calibrations were attempted for properties such as dopant concentrations, thickness, etch rate, film stress, and electrical parameters. It was found that annealed films were predicted more precisely than unannealed films. B, P, and thickness measurements yielded the most precise results by these techniques. Multivariate calibration methods applied to etch rate for annealed films and unannealed film stress provided some limited predictive ability. The detection and removal of outliers greatly improved the analysis precisions. Finally, within wafer and between wafer dopant uniformity may be responsible for degrading the precision of these analytical methods.

The proposed high-level nuclear waste repository at Yucca Mountain, Nevada will have many miles of underground openings. Because of the long-term nature of this project it is important to gain a high level of understanding of the dynamic behavior of its underground openings. The site, located on and adjacent to the Nevada Test Site (NTS), is subject to seismic loading from both natural events and underground nuclear explosions (UNEs). While ground motions from both of these seismic sources are of interest to the Yucca Mountain Project, those resulting from earthquakes are expected to be the larger of the two and, therefore, more significant in design. It would be difficult, however, to collect underground data from earthquakes because of their unpredictable nature. In contrast, UNEs have been conducted on a regular basis at the NTS and present an opportunity to obtain data useful in understanding the seismic behavior of underground openings. To this end, the Tunnel Dynamics Experiment (TDE) was fielded adjacent to a recent UNE in a pre-existing tunnel. The objective of this experiment was to document tunnel damage corresponding to measured and observed ground motions.

This report contains the full papers submitted to the Committee on the Safety of Nuclear Installations (CSNI) Workshop on Probabilistic Safety Assessment (PSA) Applications and Limitations held in Santa Fe, New Mexico, USA, on September 4 through 6, 1990. The purpose of the Workshop was to provide an avenue for discussions in the following areas: (1) current PSA result, (2) current uses of PSA, (3) views on current limitations, (4) expert opinion, and (5) low probability numbers. The papers contained herein address these issues, along with several other related topics.

This report summarizes the advanced development of the Spectrum Sciences Model 5005-TF, Single-Event Test Fixture. The Model 5005-TF uses a Californium-252 (Cf-252) fission-fragment source to test integrated circuits and other devices for the effects of single-event phenomena. Particle identification methods commonly used in high-energy physics research and nuclear engineering have been incorporated into the Model 5005-TF for estimating the particle charge, mass, and energy parameters. All single-event phenomena observed in a device under test (DUT) are correlated with an identified fission fragment, and its linear energy transfer (LET) and range in the semiconductor material of the DUT.

This report describes the features and use of the Sandia National Laboratories Assembly Test Chip Ver. 01 (ATC01). This chip contains a variety of Al conductor features which are intended for use in corrosion testing. These include triple tracks with a variety of line and gap widths, ladder structures, straight line structures, and van der Pauw sheet resistance structures. The chip is square, approximately 0.250 in. on a side, with a minimum Al feature size of 1.25 {mu}m. The various test structures on the die are described in detail and bonding layout data are given. Finally, we give an example of measurements made on ATC01 when packaged in a 40 lead CERDIP. 15 refs., 7 figs., 7 tabs.

Sub-gridding techniques enable finite-difference time-domain (FDTD) electromagnetic codes to model apertures that are much narrower than the spatial resolution of the FDTD mesh. Previous thin-slot methods have assumed that the slot walls are perfectly conducting. As the slot depth-to-width ratio becomes large, interior wall losses for realistic materials can significantly affect the coupling through the slot, and therefore these loss effects should not be neglected. This paper presents two methods for incorporating loss for walls with good, but not perfect conductivity, into the FDTD calculations. The first method modifies an FDTD equation internal to the slot to include a surface-impedance contribution. This method is appropriate for the usual FDTD thin-slot formalisms. The second method includes the losses into a half-space'' integral equation that can be used by the recently introduced Hybrid Thin-Slot Algorithm. Results based on the two methods are compared for a variety of slot parameters and wall conductivities.

The establishment of a database for the materials that are used in production Li(Si)/FeS{sub 2} thermal batteries designed at Sandia National Laboratories is described. The database is a Hewlett-Packard (HP) network type (IMAGE) designed to run on an HP3000 computer. Heavy emphasis is placed on the use of screen forms for entry, editing, and retrieval of data. Custom screen forms were used for the various materials in the battery. For the purposes of the materials database, each battery is composed of four mixes: cathode, separator, anode, and heat (pyrotechnic) powders. A consistent lot-numbering system was adopted for both the mixes and the discrete components that make up the mixes. Each serial number of a particular battery is linked to the lot numbers of the four mixes used in the battery. Each mix, in turn, is linked to the lot numbers of the discrete components that are contained within the mix. This allows traceability of each of the components used in any given serial number of a particular battery. The materials database provides the necessary traceability, as required by the Department of Energy, for the lifetime of the program associated with the battery. 3 refs., 23 figs.