Efforts to more effectively monitor nuclear explosions include the calibration of travel times along specific paths. Benchmark events are used to improve travel-time prediction by (1) improving models, (2) determining travel times empirically, or (3) using a hybrid approach. Even velocity models that are determined using geophysical analogy (i.e. models determined without the direct use of calibration data) require validation with calibration events. Ideally, the locations and origin times of calibration events would be perfectly known. However, the existing set of perfectly known events is spatially limited and many of these events occurred prior to the installation of current monitoring stations, thus limiting their usefulness. There are, however, large numbers of well (but not perfectly) located events that are spatially distributed, and many of these events may be used for calibration. Identifying the utility and limitations of the spatially distributed set of imperfect calibration data is of paramount importance to the calibration effort. In order to develop guidelines for calibration utility, we examine the uncertainty and correlation of location parameters under several network configurations that are commonly used to produce calibration-grade locations. We then map these calibration uncertainties through location procedures with network configurations that are likely in monitoring situations. …

The Inner and Outer modules of the Central Solenoid Model Coil (CSMC) were built by US and Japanese home teams in collaboration with European and Russian teams to demonstrate the feasibility of a superconducting Central Solenoid for ITER and other large tokamak reactors. The CSMC mass is about 120 t, OD is about 3.6 m and the stored energy is 640 MJ at 46 kA and peak field of 13 T. Testing of the CSMC and the CS Insert took place at the Japan Atomic Energy Research Institute (JAERI) from mid March until mid August 2000. This paper presents the main results of the tests performed.

The 3-D nonlinear toroidal gyrokinetic simulation code PG3EQ is used to study toroidal ion temperature gradient (ITG) driven turbulence--a key cause of the anomalous transport that limits tokamak plasma performance. Systematic studies of the dependence of ion thermal transport on various parameters and effects are presented, including dependence on {rvec E} x {rvec B} and toroidal velocity shear, sensitivity to the force balance in simulations with radial temperature gradient variation, and the dependences on magnetic shear and ion temperature gradient.

General physical relations connect the expected size and depth of laser damage induced craters to absorbed laser energy and to the strength of the material. In general, for small absorbers and ''instantaneous'' energy release, one expects three regions of interest. First is an inner region in which material is subjected to high pressure and temperature, pulverized and ejected. The resultant crater morphology will appear melted. A second region, outside the first, exhibits material removal due to spallation, which occurs when a shock wave is reflected at the free surface. The crater surface in this region will appear fractured. Finally, there is an outermost region where stresses are strong enough to crack material, but not to eject it. These regions are described theoretically and compared to representative observed craters in fused silica.

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A treatability test was conducted for the In Situ Redox Manipulation (ISRM) technology at the 100 D Area of the U. S. Department of Energy's Hanford Site. The target contaminant was dissolved chromate in groundwater. The ISRM technology creates a permeable subsurface treatment zone to reduce mobile chromate in groundwater to an insoluble form. The ISRM permeable treatment zone is created by reducing ferric iron to ferrous iron within the aquifer sediments, which is accomplished by injecting aqueous sodium dithionite into the aquifer and then withdrawing the reaction products. The goal of the treatability test was to create a linear ISRM barrier by injecting sodium dithionite into five wells. Well installation and site characterization activities began in spring 1997; the first dithionite injection took place in September 1997. The results of this first injection were monitored through the spring of 1998. The remaining four dithionite injections were carried out in May through July of 1998.These five injections created a reduced zone in the Hanford unconfined aquifer approximately 150 feet in length (perpendicular to groundwater flow) and 50 feet wide. The reduced zone extended over the thickness of the unconfined zone. Analysis of post-emplacement groundwater samples showed concentrations of chromate, in …

The objective of the calculation is to determine how commercial spent nuclear fuel (CSNF) stainless-steel clad assemblies are distributed over the CSNF waste packages (WPs) in the Yucca Mountain repository. The calculation defines the number of CSNF WPs that will contain stainless-steel clad assemblies, and the stainless steel content, on average, in those WPs for the Total System Performance Assessment in Support of Site Recommendation (TSPA-SR). Cladding models for TSPA-SR for stainless-steel clad assemblies differ from the models used for zirconium-clad assemblies. The information derived in this calculation helps to determine how the-cladding models are applied to WPs in TSPA-SR. The calculation addresses the WP configurations for CSNF defined in an interoffice correspondence from E.P. Stroupe to D.R. Wilkins (Stroupe 2000) and shown in Table 1.

The purpose of this calculation is to determine the surface dose rates of the short codisposal waste package (WP) of defense high-level waste (DHLW) and TRIGA (Training, Research, Isotopes, General Atomics) spent nuclear fuel (SNF). The WP contains the TRIGA SNF, in a standardized 18-in. DOE (U.S. Department of Energy) SNF canister, and five 3-m-long Savannah River Site (SRS) DHLW pour glass canisters, which surround the DOE SNF canister.

The purpose of this Analysis/Model Report (AMR) is to evaluate (by means of 2-D semianalytical and 3-D numerical models) the transport of radioactive solutes and colloids in the unsaturated zone (UZ) under ambient conditions from the potential repository horizon to the water table at Yucca Mountain (YM), Nevada. This is in accordance with the ''AMR Development Plan U0060, Radionuclide Transport Models Under Ambient Conditions'' (CRWMS M and O 1999a). This AMR supports the UZ Flow and Transport Process Model Report (PMR). This AMR documents the UZ Radionuclide Transport Model (RTM). This model considers: the transport of radionuclides through fractured tuffs; the effects of changes in the intensity and configuration of fracturing from hydrogeologic unit to unit; colloid transport; physical and retardation processes and the effects of perched water. In this AMR they document the capabilities of the UZ RTM, which can describe flow (saturated and/or unsaturated) and transport, and accounts for (a) advection, (b) molecular diffusion, (c) hydrodynamic dispersion (with full 3-D tensorial representation), (d) kinetic or equilibrium physical and/or chemical sorption (linear, Langmuir, Freundlich or combined), (e) first-order linear chemical reaction, (f) radioactive decay and tracking of daughters, (g) colloid filtration (equilibrium, kinetic or combined), and (h) colloid-assisted solute …

The purpose and scope of this analysis/model report is to analyze the degradation of commercial spent nuclear fuel (CSNF) cladding under repository conditions by the hydride-related metallurgical processes, such as delayed hydride cracking (DHC), hydride reorientation and hydrogen embrittlement, thereby providing a better understanding of the degradation process and clarifying which aspects of the process are known and which need further evaluation and investigation. The intended use is as an input to a more general analysis of cladding degradation.

The purpose of this Analysis/Model Report (AMR) is to document the Calibrated Properties Model that provides calibrated parameter sets for unsaturated zone (UZ) flow and transport process models for the Yucca Mountain Site Characterization Project (YMP). This work was performed in accordance with the ''AMR Development Plan for U0035 Calibrated Properties Model REV00. These calibrated property sets include matrix and fracture parameters for the UZ Flow and Transport Model (UZ Model), drift seepage models, drift-scale and mountain-scale coupled-processes models, and Total System Performance Assessment (TSPA) models as well as Performance Assessment (PA) and other participating national laboratories and government agencies. These process models provide the necessary framework to test conceptual hypotheses of flow and transport at different scales and predict flow and transport behavior under a variety of climatic and thermal-loading conditions.

A simulated full-scale plutonium storage cubicle with 22 horizontally positioned and heated 3013 canisters is proposed to confirm the effectiveness of natural circulation. Temperature and airflow measurements will be made for different heat generation and cubicle door configurations. Comparisons will be made to computer based thermal Hydraulic models.

This is the final report of Grant DE-FG02-92ER25119, ''Studies of Complexity in Fluids'', we have investigated turbulence, flow in granular materials, singularities in evolution of fluid surfaces and selective withdrawal fluid flows. We have studied numerical methods for dealing with complex phenomena, and done simulations on the formation of river networks. We have also studied contact-line deposition that occurs in a drying drop.

Although significant effort has been dedicated to the study of burnup-credit issues over the past decade, U.S. studies to-date have primarily focused on spent pressurized-water-reactor (PWR) fuel. The current licensing approach taken by the U.S. Department of Energy for burnup credit in transportation seeks approval for PWR fuel only. Burnup credit for boiling-water-reactor (BWR) fuel has not yet been formally sought. Burnup credit for PWR fuel was pursued first because: (1) nearly two-thirds (by mass) of the total discharged commercial spent fuel in the United States is PWR fuel, (2) it can substantially increase the fuel assembly capacity with respect to current designs for PWR storage and transportation casks, and (3) fuel depletion in PWRs is generally less complicated than fuel depletion in BWRs. However, due to international needs, the increased enrichment of modern BWR fuels, and criticality safety issues related to permanent disposal within the United States, more attention has recently focused on spent BWR fuel. Specifically, credit for fuel burnup in the criticality safety analysis for long-term disposal of spent nuclear fuel enables improved design efficiency, which, due to the large mass of fissile material that will be stored in the repository, can have substantial financial benefits. For …

The Office of Fissile Materials Disposition is responsible for disposing of inventories of surplus US weapons-usable plutonium and highly enriched uranium as well as providing, technical support for, and ultimate implementation of, efforts to obtain reciprocal disposition of surplus Russian plutonium. On January 4, 2000, the Department of Energy issued a Record of Decision to dispose of up to 50 metric tons of surplus weapons-grade plutonium using two methods. Up to 17 metric tons of surplus plutonium will be immobilized in a ceramic form, placed in cans and embedded in large canisters containing high-level vitrified waste for ultimate disposal in a geologic repository. Approximately 33 metric tons of surplus plutonium will be used to fabricate MOX fuel (mixed oxide fuel, having less than 5% plutonium-239 as the primary fissile material in a uranium-235 carrier matrix). The MOX fuel will be used to produce electricity in existing domestic commercial nuclear reactors. This paper reports the major waste-package-related, long-term disposal impacts of the two waste forms that would be used to accomplish this mission. Particular emphasis is placed on the possibility of criticality. These results are taken from a summary report published earlier this year.

If a repository were built at Yucca Mountain, it would rely on two distinct systems to prevent radioactive materials from escaping into the environment. These systems act as barriers to the movement of radionuclides. The first system involves natural barriers--the rocks, water, and climate at Yucca Mountain. The second system is comprised of an array of engineered, or man-made, barriers that give the repository defense in depth and add safety margins. These systems would work together to protect the public and the environment. The mountain's natural features present a formidable line of defense against possible movement by radionuclides. These barriers include Yucca Mountain's unique geology, the region's dry climate, and, in general, a range of enclosed water systems that should slow water that contains radioactive particles from reaching rivers or other groundwater systems. The mountain's natural barriers and planned man-made barriers should prevent most moisture from ever reaching the waste packages within a repository. Moreover, the natural barriers would slow the movement of radioactive particles that do dissolve in water. The engineering, or technological measures, that would be built into a repository at Yucca Mountain would help ensure that health and safety standards are maintained even if some components of …

This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for samples obtained from tank 241-AY-101. The purpose of this sampling event is to obtain information about the characteristics of the contents of 241-AY-101 required to satisfy Data Quality Objectives For RPP Privatization Phase I: Confirm Tank T Is An Appropriate Feed Source For High-Level Waste Feed Batch X(HLW DQO) (Nguyen 1999a), Data Quality Objectives For TWRS Privatization Phase I : Confirm Tank T Is An Appropriate Feed Source For Low-Activity Waste Feed Batch X (LAW DQO) (Nguyen 1999b), Low Activity Waste and High-Level Waste Feed Data Quality Objectives (L and H DQO) (Patello et al. 1999), and Characterization Data Needs for Development, Design, and Operation of Retrieval Equipment Developed through the Data Quality Objective Process (Equipment DQO) (Bloom 1996). Special instructions regarding support to the LAW and HLW DQOs are provided by Baldwin (1999). Push mode core samples will be obtained from risers 15G and 150 to provide sufficient material for the chemical analyses and tests required to satisfy these data quality objectives. The 222-S Laboratory will extrude core samples; composite the liquids and solids; perform chemical analyses on composite …

In this work, atomic force microscopy (AFM) has been used to identify the controlling transport mechanisms at metal/oxide interfaces and measure the corresponding diffusivities. Interfacial transport rates in our experiments are two to four orders of magnitude faster than any previously reported rates for the oxide surface. The interfacial diffusivities and the degree of interfacial anisotropy depend on the oxygen activity of the system. Atomic transport at metal/oxide interfaces plays a defining role in many technological processes, and these experiments provide fundamental data for the formulation of the atomic theory needed to explain many of the observed phenomena.

This analysis identifies the DST Subsystem functions for storing, transferring, receiving, and preparing waste in support of the Waste Feed Delivery mission.

The compatibility program described in this document formalizes the process for determining waste compatibility. The primary goal of the program is to ensure that sufficient controls are in place to prevent the formation of incompatible mixtures during future operations. The process described involves characterizing waste, comparing characteristics with criteria, resolving potential incompatibilities and documenting the process.

This document provides the calculations used to determine the error of safety class signals used for the CVD process These errors are used with the Parameter limits to arrive at the initial set point. The Safety Class Instrumentation and Control (SCIC) system provides active detection and response to process anomalies that, if unmitigated would result in a safety event. Specifically actuation of the SCIC system includes two portions. The portion which isolates the MCO and initiates the safety-class helium (SCHe) purge, and the portion which detects and stops excessive heat input to the MCO on high tempered water MCO inlet temperature. For the MCO isolation and purge the SCIC receives signals from MCO pressure (both positive pressure and vacuum) helium flow rate, bay high temperature switches, seismic trips and time under vacuum trips.

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The management plan developed to characterize the K Basin spent nuclear fuel (SNF) and sludge was originally developed for Westinghouse Hanford Company and Pacific Northwest National Laboratory to work together on a program to provide characterization data to support removal, conditioning, and subsequent dry storage of the SNF stored at the Hanford K Basins. The plan also addressed necessary characterization for the removal, transport, and storage of the sludge from the Hanford K Basins. This plan was revised in 1999 (i.e., Revision 2) to incorporate actions necessary to respond to the deficiencies revealed as the result of Quality Assurance surveillances and audits in 1999 with respect to the fuel characterization activities. Revision 3 to this Program Management Plan responds to a Worker Assessment resolution determined in Fical Year 2000. This revision includes an update to current organizational structures and other revisions needed to keep this management plan consistent with the current project scope. The plan continues to address both the SNF and the sludge accumulated at K Basins. Most activities for the characterization of the SNF have been completed. Data validation, Office of Civilian Radioactive Waste Management (OCRWM) document reviews, and OCRWM data qualification are the remaining SNF characterization activities. …

The goal of the Savannah River Salt Waste Processing Program (SPP) is to evaluate and select the most effective technology for the treatment of the high-level waste salt solutions currently being stored in underground storage tanks at the U.S. Department of Energy Savannah River Site (SRS) in Aiken, South Carolina. One of the three technologies currently being developed for this application is the Small-Tank Tetraphenylborate Process (STTP). This process uses sodium tetraphenylborate (NaTPB) to precipitate and remove radioactive cesium from the waste and monosodium titanate (MST) to sorb and remove radioactive strontium and actinides. Oak Ridge National Laboratory is demonstrating this process at the 1:4000 scale using a 20-L capacity continuous-flow stirred-tank reactor (CSTR) system. Since March 1999, three operating campaigns of the 20-L CSTR have been conducted. The ultimate goal is to verify that this process, under certain extremes of operating conditions, can meet the minimum treatment criteria necessary for processing and disposal at the Savannah River Saltstone Facility. The waste acceptance criteria (WAC) for {sup 137}Cs, {sup 90}Sr, and total actinides are <40 nCi/g, <40 nCi/g, and <18 nCi/g, respectively. However, to allow for changes in process conditions, SPP is seeking a level of treatment that is about …