Publications

Distributed denial of service (DoS) attacks on cyber-resources are complex problems that are difficult to completely define, characterize, and mitigate. We recognize the process-nature of DoS attacks and view them from multiple perspectives. Identification of opportunities for mitigation and further research may result from this attempt to characterize the DoS problem space. We examine DoS attacks from the point of view of (1) a high-level that establishes common terminology and a framework for discussing the DoS process, (2) layers of the communication stack, from attack origination to the victim of the attack, (3) specific network and computer elements, and (4) attack manifestations. We also examine DoS issues associated with wireless communications. Using this collection of views, one begins to see the DoS problem in a holistic way that may lead to improved understanding, new mitigation strategies, and fruitful research.

The Russia-US joint program on the safe management of nuclear materials was initiated to address common technical issues confronting the US and Russia in the management of excess weapons grade nuclear materials. The program was initiated after the 1993 Tomsk-7 accident. This paper provides an update on program activities since 1996. The Fourth US Russia Nuclear Materials Safety Management Workshop was conducted in March 1997. In addition, a number of contracts with Russian Institutes have been placed by Lawrence Livermore National Laboratory (LLNL) and Sandia National Laboratories (SNL). These contracts support research related to the safe disposition of excess plutonium (Pu) and highly enriched uranium (HEU). Topics investigated by Russian scientists under contracts with SNL and LLNL include accident consequence studies, the safety of anion exchange processes, underground isolation of nuclear materials, and the development of materials for the immobilization of excess weapons Pu.

This document describes the DOSFAC2 code, which is used for generating dose-to-source conversion factors for the MACCS2 code. DOSFAC2 is a revised and updated version of the DOSFAC code that was distributed with version 1.5.11 of the MACCS code. included are (1) an overview and background of DOSFAC2, (2) a summary of two new functional capabilities, and (3) a user`s guide. 20 refs., 5 tabs.

MACCS2 represents a major enhancement of the capabilities of its predecessor MACCS, the MELCOR Accident Consequence Code System. MACCS, released in 1987, was developed to estimate the potential impacts to the surrounding public of severe accidents at nuclear power plants. The principal phenomena considered in MACCS/MACCS2 are atmospheric transport and deposition under time-variant meteorology, short-term and long-term mitigative actions and exposure pathways, deterministic and stochastic health effects, and economic costs. MACCS2 was developed as a general-purpose analytical tool applicable to diverse reactor and nonreactor facilities. The MACCS2 package includes three primary enhancements: (1) a more flexible emergency response model, (2) an expanded library of radionuclides, and (3) a semidynamic food-chain model. In addition, errors that had been identified in MACCS version1.5.11.1 were corrected, including an error that prevented the code from providing intermediate-phase results. MACCS2 version 1.10 beta test was released to the beta-test group in May, 1995. In addition, the University of New Mexico (UNM) has completed an independent verification study of the code package. Since the beta-test release of MACCS2 version 1.10, a number of minor errors have been identified and corrected, and a number of enhancements have been added to the code package. The code enhancements added since the beta-test release of version 1.10 include: (1) an option to allow the user to input the {sigma}{sub y} and {sigma}{sub z} plume expansion parameters in a table-lookup form for incremental downwind distances, (2) an option to define different initial dimensions for up to four segments of a release, (3) an enhancement to the COMIDA2 food-chain model preprocessor to allow the user to supply externally calculated tables of tritium food-chain dose per unit deposition on farmland to support analyses of tritium releases, and (4) the capability to calculate direction-dependent doses.

CHEM{_}MACCS has been developed from the radiological accident consequence code, MACCS, to perform probabilistic calculations of potential off-site consequences of the accidental atmospheric release of hazardous chemicals. The principal phenomena considered in CHEM{_}MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways, and early and latent health effects. CHEM{_}MACCS provides the following capabilities: (1) statistical weather sampling data (8,760 hourly data points per year), (2) population dose and health effect risk calculations based on site-specific population data, (3) health effects calculations including the consideration of potential site specific mitigative actions (evacuation and shielding), and (4) modeling of multiple release segments. Three different sample problems are contained in this report to show how to use CHEM{_}MACCS. Three test problems are run to compare CHEM{_}MACCS and D2PC. The doses versus the downwind centerline distances from the source for the given doses are in very close agreement.

The MELCOR Accident Consequence Code System (MACCS), publicly distributed since 1987, was developed to estimate the potential impacts to the surrounding public of severe accidents at nuclear power plants. The principal phenomena considered in MACCS are atmospheric transport and deposition under time-variant meteorology, short-term and long-term mitigative actions and exposure pathways, deterministic and stochastic health effects, and economic costs of mitigative actions. At this time, no other publicly available code in the US offers all these capabilities. MACCS2 represents a major enhancement of the capabilities of its predecessor MACCS. MACCS2 was developed as a general-purpose analytical tool applicable to diverse reactor and nonreactor Department of Energy (DOE) facilities. The MACCS2 package includes three primary enhancements: (1) a more flexible emergency response model, (2) an expanded library of radionuclides, and (3) a semidynamic food-chain model. The new code features allow detailed evaluations of risks to workers at nearby facilities on large DOE reservations and allow the user to assess the potential impacts of over 700 radionuclides that cannot be considered with MACCS.

MACCS was developed at Sandia National Laboratories (SNL) under U.S. Nuclear Regulatory Commission (NRC) sponsorship to estimate the offsite consequences of potential severe accidents at nuclear power plants (NPPs). MACCS was publicly released in 1990. MACCS was developed to support the NRC`s probabilistic safety assessment (PSA) efforts. PSA techniques can provide a measure of the risk of reactor operation. PSAs are generally divided into three levels. Level one efforts identify potential plant damage states that lead to core damage and the associated probabilities, level two models damage progression and containment strength for establishing fission-product release categories, and level three efforts evaluate potential off-site consequences of radiological releases and the probabilities associated with the consequences. MACCS was designed as a tool for level three PSA analysis. MACCS performs probabilistic health and economic consequence assessments of hypothetical accidental releases of radioactive material from NPPs. MACCS includes models for atmospheric dispersion and transport, wet and dry deposition, the probabilistic treatment of meteorology, environmental transfer, countermeasure strategies, dosimetry, health effects, and economic impacts. The computer systems MACCS is designed to run on are the 386/486 PC, VAX/VMS, E3M RISC S/6000, Sun SPARC, and Cray UNICOS. This paper provides an overview of MACCS, reviews some of the applications of MACCS, international collaborations which have involved MACCS, current developmental efforts, and future directions.

HEISHI is a Fortran computer model designed to aid in analysis, prediction, and optimization of fuel characteristics for use in Space Nuclear Thermal Propulsion (SNTP). Calculational results include fission product release rate, fuel failure fraction, mode of fuel failure, stress-strain state, and fuel material morphology. HEISHI contains models for decay chain calculations of retained and released fission products, based on an input power history and release coefficients. Decay chain parameters such as direct fission yield, decay rates, and branching fractions are obtained from a database. HEISHI also contains models for stress-strain behavior of multilayered fuel particles with creep and differential thermal expansion effects, transient particle temperature profile, grain growth, and fuel particle failure fraction. Grain growth is treated as a function of temperature; the failure fraction depends on the coating tensile strength, which in turn is a function of grain size. The HEISHI code is intended for use in analysis of coated fuel particles for use in particle bed reactors; however, much of the code is geometry-independent and applicable to fuel geometries other than spherical.

The NRC has proposed revisions to 10 CFR 100 which include the codification of nuclear reactor site population density limits to 500 people per square mile, at the siting stage, averaged over any radial distance out to 30 miles, and 1,000 people per square mile within the 40-year lifetime of a nuclear plant. This study examined whether there are less restrictive alternative population density and/or distribution criteria which would provide equivalent or better protection to human health in the unlikely event of a nuclear accident. This study did not attempt to directly address the issue of actual population density limits because there are no US risk standards established for the evaluation of population density limits. Calculations were performed using source terms for both a current generation light water reactor (LWR) and an advanced light water reactor (ALWR) design. The results of this study suggest that measures which address the distribution of the population density, including emergency response conditions, could result in lower average individual risks to the public than the proposed guidelines that require controlling average population density. Studies also indicate that an exclusion zone size, determined by emergency response conditions and reactor design (power level and safety features), would better serve to protect public health than a rigid standard applied to all sites.

The degree of oxidation of drops of aluminum metal was investigated parametrically for a range of melt diameters, relative melt-water velocities, melt temperatures, water temperatures, and ambient pressures using a combined film boiling-metal oxidation model. The model predictions of degree of oxidation were then compared to small-scale experiments involving molten drops of aluminum metal. The conclusions were, first, that for the range of melt temperatures and diameters considered, if an oxide layer forms and blankets the molten aluminum, then no significant oxidation occurs, in agreement with the results of experiments performed under quiescent, steady-state conditions. Second, comparing model results to data from single drop fragmentation experiments indicates that under the transient conditions occurring during rapid fragmentation, the oxide layer is disturbed and oxidation rates are limited primarily by the amount of steam available at the melt surface. Third, for a range of parameters, the heat gain in the melt drop from the oxidation reaction can exceed the heat loss to the surroundings, resulting in thermal runaway and ignition of the melt. This effect is observed experimentally as a threshold temperature effect, predicted to be about 1100 K for the initial single drop study, and between 1500 K and 1600 K for the single drop experiments. 7 refs., 2 tabs.

The phenomenon of molten fuel-coolant interaction (FCI) is of considerable interest in many industrial processes where hot molten material may come in contact with water, including the pulp and paper, aluminum, steel, and nuclear power industries. The nature of the FCIs can range from mild film boiling, through energetic boiling, up to a violent vapor explosion. In the nuclear power industry, FCIs are of interest because of their possible consequences during hypothetical light water reactor core meltdown accidents. These interactions may occur under a variety of conditions either within the reactor vessel or in the reactor cavity. The IFCI computer code is being developed to investigate the FCI problem at large scale using a two-dimensional, four-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. The hydrodynamic method and physical models used in IFCI are discussed. Results from a test problem simulating a generic pouring mode experiment are presented. 39 refs., 10 figs., 1 tab.