Publications

The validity of scale model impact evaluation of the SST Transportation System is acceptable based on Dimensional Analysis (Buckingham Pi Theorem) and the work of numerous programs that have evaluated the agreement among dimensional analysis, several different reduced-size models and full-scale impact test data. Excellent accuracy has been demonstrated between scale models and full-scale impact data when collected in conformance with the Buckingham Pi Theorem. 20 refs., 4 figs.

A five-step procedure was used in the 1990 performance simulations to construct probability distributions of the uncertain variables appearing in the mathematical models used to simulate the Waste Isolation Pilot Plant's (WIPP's) performance. This procedure provides a consistent approach to the construction of probability distributions in cases where empirical data concerning a variable are sparse or absent and minimizes the amount of spurious information that is often introduced into a distribution by assumptions of nonspecialist. The procedure gives first priority to the professional judgment of subject-matter experts and emphasizes the use of site-specific empirical data for the construction of the probability distributions when such data are available. In the absence of sufficient empirical data, the procedure employs the Maximum Entropy Formalism and the subject-matter experts' subjective estimates of the parameters of the distribution to construct a distribution that can be used in a performance simulation. 23 refs., 4 figs., 1 tab.

This report documents the data available as of August 1990 and used by the Performance Assessment Division of Sandia National Laboratories in its December 1990 preliminary performance assessment of the Waste Isolation Pilot Plant (WIPP). Parameter values are presented in table form for the geologic subsystem, engineered barriers, borehole flow properties, climate variability, and intrusion characteristics. Sources for the data and a brief discussion of each parameter are provided. 101 refs., 72 figs., 21 tabs.

A theoretical model was developed for the minimum charge to trigger a gaseous detonation in spherical geometry as a generalization of the Zeldovich model. Careful comparisons were made between the theoretical predictions and experimental data on the minimum charge to trigger detonations in propane-air mixtures. The predictions are an order of magnitude too high, and there is no apparent resolution to the discrepancy. A dynamic model, which takes into account the experimentally observed oscillations in the detonation zone, may be necessary for reliable predictions. 27 refs., 9 figs.

The Salado Formation is a thick evaporite sequence located in the Permian Delaware Basin of southeastern New Mexico. This study focuses on the intense diagenetic alteration that has affected the small amounts of clay, feldspar, and quartz washed into the basin during salt deposition. These changes are of more than academic interest since this formation also houses the WIPP (Waste Isolation Pilot Plant). Site characterization concerns warrant compiling a detailed data base describing the clays in and around the facility horizon. An extensive sampling effort was undertaken to address these programmatic issues as well as to provide additional insight regarding diagenetic mechanisms in the Salado. Seventy-five samples were collected from argillaceous partings in halite at the stratigraphic level of the Waste Isolation Pilot Plant (WIPP). These were compared with twenty-eight samples from cores of the Vaca Triste member of the Salado, a thin clastic unit at the top of the McNutt potash zone, and with a clay-rich sample from the lower contact of the Culebra Dolomite (in the overlying Rustler Formation). These settings were compared to assess the influence of differences in brine chemistry (i.e., halite and potash facies, normal to hypersaline marine conditions) and sediment composition (clays, sandy silt, dolomitized limestone) on diagenetic processes. 44 refs., 11 figs., 5 tabs.

The Waste Isolation Pilot Plant (WIPP) is planned as the first mined geologic repository for transuranic (TRU) wastes generated by defense programs of the United States Department of Energy (DOE). Before disposing of waste at the WIPP, the DOE must evaluate compliance with the United states Environmental Protection Agency's (EPA) Standard, Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes (40 CFR Part 191, US EPA, 1985). Sandia National Laboratories (SNL) is evaluating long-term performance against criteria in Subpart B of the Standard. Performance assessment'' as used in this report includes analyses for the Containment Requirements ({section} 191.13(a)) and the Individual Protection Requirements ({section} 191.15). Because proving predictions about future human actions or natural events is not possible, the EPA expects compliance to be determined on the basis of specified quantitative analyses and informed, qualitative judgment. The goal of the WIPP performance-assessment team at SNL is to provide as detailed and thorough a basis as practical for the quantitative aspects of that decision. This report summarizes SNL's late-1990 understanding of the WIPP Project's ability to evaluate compliance with Subpart B. 245 refs., 88 figs., 23 tabs.

Calculations were performed with CTH (a finite difference hydrodynamics code) to evaluate computational capabilities for predicting residual projectile length and position in high velocity penetration events. The calculations simulated selected tests in a set of armor penetration experiments conducted and reported by Los Alamos National Laboratory. The tests and simulations involved penetration of armor ranging from 6.4 to 50.8 mm in thickness by long rod projectiles over a range of velocities from 1.0 to 1.29 km/sec. Comparisons are made between the calculated and measured final projectile lengths and positions, and the sensitivity of the predicted results to target and projectile property variations is investigated. 8 refs., 11 figs., 8 tabs.

This study supports the Waste Isolation Pilot Plant (WIPP) Final Supplemental Environmental Impact Statement and has two main objectives. First, it describes current ideas about the characteristics and potential impacts of the disturbed-rock zone (DRZ) known to develop with time around excavations at the WIPP horizon. Second, it presents new calculations of radionuclide migration within and from the WIPP repository for steady-state undisturbed conditions and for two cases that consider human intrusion into the repository. At the WIPP, the presence of a DRZ has been confirmed by geophysical studies, gas-flow tests, and direct observations. The DRZ will allow gas or brine from waste-emplacement panels to bypass panel seals and flow into adjacent portions of the underground workings unless preventive measures are taken. Revised calculations of the undisturbed performance of the repository indicate that no radionuclides will be released into the Culebra Dolomite within the regulatory period of 10,000 years. The human-intrusion calculations included here assume a connection between the WIPP repository, an occurrence of pressurized brine within the underlying Castile Formation, and the overlying Culebra Dolomite. 61 refs., 40 figs., 16 tabs.

The problem of calculating the vibrations of rotating structures has challenged analysts since the observation that use of traditional modal coordinates in such problems leads to the prediction of instability involving infinite deformation when rotation rates exceed the first natural frequency. Much recent published work on beams has shown that such predictions are artifacts of incorporating incomplete kinematics into the analysis, but that work addresses analysis of only simple structures such as individual beams and plates. The authors present a new approach to analyzing rotating flexible structures that applies to the rotation of general linear (unjointed) structures, using a system of nonlinearly coupled deformation modes. This technique is called a Method of Quadratic Modes. 37 refs., 5 figs., 1 tab.

The geochemical modeling codes EQ3NR/EQ6 were used to model the interaction of cementitious materials with ground water from the Yucca Mountain proposed nuclear waste repository site in Nevada. This paper presents a preliminary estimate of the compositional changes caused by these interactions in the ground water and in the cement-based compounds proposed for use as sealing and shaft liner materials at the Yucca Mountain site. The geochemical speciation/solubility/reaction path codes EQ3NR/EQ6 were used to model the interaction of cementitious materials and water. Interaction of water with a cementitious material will result in dissolution of certain cement phases and changes in the water chemistry. These changes in the water chemistry may further lead to the precipitation of minerals either in the concrete or in the surrounding tuff at the Yucca Mountain Site (YMS). As part of a larger scoping study, a range of water, cement, and tuff compositions, temperatures, and reaction path modes were used. This paper presents a subset of that study by considering the interaction of three different cement formulations at 25{degree}C with J-13 water using the ``closed`` reaction path mode. This subset was chosen as a base case to answer important questions in selecting the compositions of cementitious materials for use in the proposed repository. 8 refs., 1 fig., 3 tabs.

American Samoa has no indigenous fossil fuels and is almost totally dependent for energy on seaborne petroleum. However, the seven Pacific Islands located at 14 degrees south latitude that constitute American Samoa have a wide variety of renewable resources with the potential for substituting for imported oil. Included as possible renewable energy conversion technologies are solar thermal, photovoltaics, wind, geothermal, ocean thermal, and waste-to-energy recovery. This report evaluates the potential of each of these renewable energy alternatives and establishes recommended priorities for their development in American Samoa. Rough cost estimates are also included. Although renewable energy planning is highly site specific, information in this report should find some general application to other tropical insular areas.

In this report, we provide a detailed discussion of methodology of predicting cable degradation versus dose rate, temperature, and exposure time and its application to data obtained on a number of additional nuclear power plant cable insulation (a hypalon, a silicon rubber and two ethylenetetrafluoroethylenes) and jacket (a hypalon) materials. We then show that the predicted, low-dose-rate results for our materials are in excellent agreement with long-term (7 to 9 years), low dose-rate results recently obtained for the same material types actually aged under nuclear power plant conditions. Based on a combination of the modelling and long-term results, we find indications of reasonably similar degradation responses among several different commercial formulations for each of the following generic'' materials: hypalon, ethylenetetrafluoroethylene, silicone rubber and PVC. If such generic'' behavior can be further substantiated through modelling and long-term results on additional formulations, predictions of cable life for other commercial materials of the same generic types would be greatly facilitated. Finally, to aid utilities in their cable life extension decisions, we utilize our modelling results to generate lifetime prediction curves for the materials modelled to data. These curves plot expected material lifetime versus dose rate and temperature down to the levels of interest to nuclear power plant aging. 18 refs., 30 figs., 3 tabs.

PRONTO 2D and PRONTO 3D are two- and three-dimensional transient solid dynamics codes for analyzing large deformations of highly nonlinear materials subjected to high strain rates. This newsletter is issued to document changes to these codes. As of this writing, the latest version of PRONTO 2D is Version 4.5.6, and the latest version of PRONTO 3D is Version 4.5.6. This update of the two codes discusses two major modifications to the numerical formulations, three new constitutive models, and the additions and improvements of contact surfaces. Changes in file formats, other miscellaneous revisions, and the availability of PRONTO 2D and PRONTO 3D are also discussed. In addition, updated commands for PRONTO 2D are provided in Appendix A of this newsletter. 29 refs., 12 figs., 2 tabs.

This report analyzes the effects of finite-precision arithmetic on discrete Fourier transforms (DFTs) calculated using the chirp-z transform algorithm. An introduction to the chirp-z transform is given together with a description of how the chirp-z transform is implemented in hardware. Equations for the effects of chirp rate errors, starting frequency errors, and starting phase errors on the frequency spectrum of the chirp-z transform are derived. Finally, the maximum possible errors in the chirp rate, the starting frequencies, and starting phases are calculated and used to compute the worst case effects on the amplitude and phase spectrums of the chirp-z transform. 1 ref., 6 figs.

Numerous gas flow measurements have been made at the Waste Isolation Pilot Plant (WIPP) Facility horizon during 1988. All tests have been pressure decay or constant pressure tests from single boreholes drilled from the underground excavations. The test fluid has been nitrogen. The data have been interpreted as permeabilities and porosities by means of a transient numerical solution method. A closed-form steady-state approximation provides a reasonable order-of-magnitude permeability estimate. The effective resolution of the measurement system is less than 10{sup {minus}20} m{sup 2}. Results indicate that beyond 1 to 5 m from an excavation, the gas flow is very small and the corresponding permeability is below the system resolution. Within the first meter of an excavation, the interpreted permeabilities can be 5 orders of magnitude greater than the undisturbed or far-field permeability. The interpreted permeabilities in the region between the undisturbed region and the first meter from an excavation are in the range of 10{sup {minus}16} to 10{sup {minus}20} m{sup 2}. Measurable gas flow occurs to a greater depth into the roof above WIPP excavations of different sizes and ages than into the ribs and floor. The gas flows into the formation surrounding the smallest excavation tested are consistently lower than those at similar locations surrounding larger excavations of comparable age. Gas flow measured in the interbed layers near the WIPP excavations is highly variable. Generally, immediately above and below excavations, relatively large gas flow is measured in the interbed layers. These results are consistent with previous measurements and indicate a limited disturbed zone surrounding WIPP excavations. 31 refs., 99 figs., 5 tabs.

A workstation manufactured by International Business Machines Corporation (IBM) was loaned to the Simulation Technology Research Division for evaluation. We have found that these new UNIX workstations from IBM have superior cost to performance ratios compared to the CRAY supercomputers and Digital's VAX machines. Our appraisal of this workstation included floating-point performance, system and environment functionality, and cost effectiveness. Our assessment was based on a suite of radiation transport codes developed at Sandia that constitute the bulk of our division's computing workload. In this report, we also discuss our experience with features that are unique to this machine such as the AIX operating system and the XLF Fortran Compiler. The interoperability of the RS/6000 workstation with Sandia's network of CRAYs and VAXs was also assessed.

A pilot plant project was undertaken to develop a central refrigeration compressor station capable of providing the necessary cooling to a network of nine independently-controlled environmental test chambers operating at temperatures of {minus}85{degree}F to 350{degree}F. Design features of the central two-stage (cascade) vapor compression refrigeration system are described. Computer control of the central refrigeration station is a major contribution to the improved efficiency of the overall system. A second computer-control system was developed to perform the task of environmental chamber control, test management, and chamber performance monitoring. Data on performance of the Climatic Central Refrigeration System (CCRS) are presented. 7 refs., 18 figs.

A complete external events probabilistic risk assessment has been performed for the N-Reactor power plant, making full use of all insights gained during the past ten years' developments in risk assessment methodologies. A detailed screening analysis was performed which showed that all external events had negligible contribution to core damage frequency except fires, seismic events, and external flooding. A limited scope analysis of the external flooding risk indicated that it is not a major risk contributor. Detailed analyses of the fire and seismic risks resulted in total (mean) core damage frequencies of 1.96E-5 and 4.60E-05 per reactor year, respectively. Detailed uncertainty analyses were performed for both fire and seismic risks. These results show that the core damage frequency profile for these events is comparable to that found for existing commercial power plants if proposed fixes are completed as part of the restart program. 108 refs., 85 figs., 80 tabs.

Automated entry control has become an increasingly important issue at facilities where budget constraints are limiting options for manned entry control points. Three questions are immediately raised when automated entry control is considered: What hardware is available How much does it cost How effective is it in maintaining security Ongoing work at Sandia National Labs is attempting to answer these questions and establish a data base for use by facility security managers working the problem of how to maintain security on a limited budget. 14 refs.

A wide variety of physical phenomena arising within many scientific disciplines can be described by systems of coupled partial differential equations (PDEs). The numerical approximation of these PDEs often involves the solution of a system of algebraic equations (possibly nonlinear) which are typically large, sparse and nonsymmetric. The increasing computational demands required by the solution of such complex scientific applications has motivated the current direction toward large-scale parallel computers. We, therefore, consider solution techniques of representative systems of equations on large scale MIMD machines. Our primary emphasis in this study is the evaluation of iterative methods for the solution of nonsymmetric systems. In particular, we discuss two Krylov subspace methods, the conjugate gradient squared algorithm (CGS) and the generalized minimum residual method (GMRES), along with the multigrid algorithm (MG) on massively parallel MIMD architectures. The focus of this evaluation considers the performance of various algorithm and implementation variations over a broad selection of problems using a parallel machine.

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.

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.

Particle Impact Noise Detection (PIND) testing is widely used to test electronic devices for the presence of conductive particles which can cause catastrophic failure. This paper develops a statistical model based on the rate of particles contaminating the part, the rate of particles induced by the test vibration, the escape rate, and the false alarm rate. Based on data from a large number of PIND tests for a canned transistor, the model is shown to fit the observed results closely. Knowledge of the parameters for which this fit is made is important in evaluating the effectiveness of the PIND test procedure and for developing background judgment about the performance of the PIND test. Furthermore, by varying the input parameters to the model, the resulting yield, failure rate and percent fallout can be examined and used to plan and implement PIND test programs.

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.

The success of probabilistic risk assessment (PRA) has led to numerous improvements to the safety of commercial nuclear power plants. Those very successes, however, have led to situation where the events that PRAs have traditionally modeled are extremely rare. As a result, current PRAs have the potential to be misleading because events previously considered to be unimportant and, thus, not included in the scope of the analyses may now be dominant. This paper provides an assessment of the current status of accident sequence analysis methods and identifies particular limitations that should be addressed in future studies.

Level 2 PRA analysis can be extremely costly. There are several options that can be used to avoid the high expenditures of time and money to obtain Level 2 insights. With reduced expenditure comes reduced insight into the behavior of specific parameters that contribute to containment failure and large source terms. If the primary objective of a study is quantitative, and does not require insights concerning the complex interactions among physical parameters and system parameters, a reduced scope method can be acceptable. If primary objectives require the understanding of the relative importance of specific physical processes and the complex interactions among the processes, reduced scope methods should not be used.

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.

A method which is capable of efficient calculation of the axisymmetric flow field produced by a large system of ring vortices is presented in this report. The system of ring vortices can, in turn, be used to model body surfaces and wakes in incompressible unsteady axisymmetric flow fields. This method takes advantage of source point and field point series expansions which enables one to make calculations for interactions between groups of vortices which are in well separated spatial domains rather than having to consider interactions between every pair of vortices. In this work, series expansions for the stream function of the ring vortex system are obtained. Such expansions explicitly contain the radial and axial velocity components. A Fortran computer code RSOLV has been written to execute the fast solution technique to calculate the stream function and the axial and radial velocity components at points in the flow field. Test cases have been run to optimize the code and to benchmark the truncation errors and CPU time savings associated with the method. Non-dimensional truncation errors for the stream function and total velocity field are on the order of 5 {times} 10{sup {minus}5} and 3 {times} 10{sup {minus}3} respectively. Single precision accuracy produces errors in these quantities up to about 1 {times} 10{sup {minus}5}. For 100 vortices in the field, there is virtually no CPU time savings with the fast solver. For 10,000 vortices in the flow, the fast solver obtains solutions in about 1% to 3% of the time required for the direct solution technique. Simulations of vortices with square and circular cores were run in order to obtain expressions for the self-induced velocities of such vortices. 8 refs., 26 figs.

A balloon-borne integrating nephelometer has been successfully developed and flown by Sandia National Laboratories and Radiance Research. This report details instrument design, calibration and data conversion procedure. Free and tethered balloon transport and telemetry systems are described. Data taken during March 1989 South-Central New Mexico free flight ascents are presented as vertical profiles of atmospheric particle scattering coefficient, temperature and balloon heading. Data taken during December 1989 Albuquerque, New Mexico tethered flights are also presented as vertical profiles. Data analysis shows superior instrument performance. 5 refs., 22 figs.

This report describes the computer program used at the Tonopah Test Range to maintain the explosive inventory. The program, which uses dBASE III or dBASE III Plus and runs on an IBM PC or compatible, has the capabilities to update (add or subtract) items, edit or delete, append, and generate various reports.