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Optimization of Hydride Rim Formation in Unirradiated Zr 4 Cladding

Description: The purpose of this work is to build on the results reported in the M2 milestone M2FT 13PN0805051, document number FCRD-USED-2013-000151 (Hanson, 2013). In that work, it was demonstrated that unirradiated samples of zircaloy-4 cladding could be pre-hydrided at temperatures below 400°C in pure hydrogen gas and that the growth of hydrides on the surface could be controlled by changing the surface condition of the samples and form a desired hydride rim on the outside diameter of the cladding. The work performed at Pacific Northwest National Laboratory since the issuing of the M2 milestone has focused its efforts to optimize the formation of a hydride rim on available zircaloy-4 cladding samples by controlling temperature variation and gas flow control during pre-hydriding treatments. Surface conditioning of the outside surface was also examined as a variable. The results of test indicate that much of the variability in the hydride thickness is due to temperature variation occurring in the furnaces as well as how hydrogen gas flows across the sample surface. Efforts to examine other alloys, gas concentrations, and different surface conditioning plan to be pursed in the next FY as more cladding samples become available
Date: September 30, 2013
Creator: Shimskey, Rick W.; Hanson, Brady D. & MacFarlan, Paul J.
Partner: UNT Libraries Government Documents Department

Hanford`s progress toward dry interim storage of K basin`s spent fuel

Description: This paper highlights the progress made toward removing the U.S. Department of Energy`s (DOE) approximately 2, 100 metric tons of metallic spent nuclear fuel from two outdated K Basins on the banks of the Columbia River and placing it in safe, economic interim dry storage beginning in December 1997. A new way of doing business at the Hanford Site and within DOE is being used to achieve the fast-track schedule, , cost savings, and public cooperation needed for success. In February 1994, the Spent Nuclear Fuel (SNF) Project was formed to solve serious safety and environmental problems associated with corroding metallic spent fuel stored in 1950`s vintage, leak-prone, water- filled concrete basins located within 365 meters (400 yards) of the last remaining unspoiled section of the Columbia River. Working together, the integrated project team focused on quickly getting the fuel out of the basins and into safe, dry storage. The team involved the public, government, regulators, and other stakeholders and forged a common understanding. The DOE transferred authority to the field to shorten approval times, and Site contractors reengineered processes to improve efficiency. Within nine months of creating the project, a plan was recommended to the DOE. It was approved on February 14, 1995. Further refinement, during the following six months, shortened the schedule even more and reduced costs by $350 million. The SNF Project is on a fast track. The K Basins Environmental Impact Statement was completed in only 11 months for only $1.3 million. Fuel and sludge samples were obtained from both basins and were sent to the laboratory for characterization and testing. The partially constructed Canister Storage Building (CSB), selected as the fuel storage facility, was redesigned, and construction was restarted saving over $17 million and cutting a year off the project schedule. With fuel removal beginning ...
Date: May 9, 1996
Creator: Culley, G.E., Westinghouse Hanford
Partner: UNT Libraries Government Documents Department

Viability of Existing INL Facilities for Dry Storage Cask Handling

Description: This report evaluates existing capabilities at the INL to determine if a practical and cost effective method could be developed for opening and handling full-sized dry storage casks. The Idaho Nuclear Technology and Engineering Center (INTEC) CPP-603, Irradiated Spent Fuel Storage Facility, provides the infrastructure to support handling and examining casks and their contents. Based on a reasonable set of assumptions, it is possible to receive, open, inspect, remove samples, close, and reseal large bolted-lid dry storage casks at the INL. The capability can also be used to open and inspect casks that were last examined at the TAN Hot Shop over ten years ago. The Castor V/21 and REA-2023 casks can provide additional confirmatory information regarding the extended performance of low-burnup (<45 GWD/MTU) used nuclear fuel. Once a dry storage cask is opened inside CPP-603, used fuel retrieved from the cask can be packaged in a shipping cask, and sent to a laboratory for testing. Testing at the INL’s Materials and Fuels Complex (MFC) can occur starting with shipment of samples from CPP-603 over an on-site road, avoiding the need to use public highways. This reduces cost and reduces the risk to the public. The full suite of characterization methods needed to establish the condition of the fuel exists and MFC. Many other testing capabilities also exist at MFC, but when those capabilities are not adequate, samples can be prepared and shipped to other laboratories for testing. This report discusses how the casks would be handled, what work needs to be done to ready the facilities/capabilities, and what the work will cost.
Date: April 1, 2013
Creator: Bohachek, Randy; Park, Charles; Wallace, Bruce; Winston, Phil & Marschman, Steve
Partner: UNT Libraries Government Documents Department

The Effect of Weld Residual Stress on Life of Used Nuclear Fuel Dry Storage Canisters

Description: With the elimination of Yucca Mountain as the long-term storage facility for spent nuclear fuel in the United States, a number of other storage options are being explored. Currently, used fuel is stored in dry-storage cask systems constructed of steel and concrete. It is likely that used fuel will continue to be stored at existing open-air storage sites for up to 100 years. This raises the possibility that the storage casks will be exposed to a salt-containing environment for the duration of their time in interim storage. Austenitic stainless steels, which are used to construct the canisters, are susceptible to stress corrosion cracking (SCC) in chloride-containing environments if a continuous aqueous film can be maintained on the surface and the material is under stress. Because steel sensitization in the canister welds is typically avoided by avoiding post-weld heat treatments, high residual stresses are present in the welds. While the environment history will play a key role in establishing the chemical conditions for cracking, weld residual stresses will have a strong influence on both crack initiation and propagation. It is often assumed for modeling purposes that weld residual stresses are tensile, high and constant through the weld. However, due to the strong dependence of crack growth rate on stress, this assumption may be overly conservative. In particular, the residual stresses become negative (compressive) at certain points in the weld. The ultimate goal of this research project is to develop a probabilistic model with quantified uncertainties for SCC failure in the dry storage casks. In this paper, the results of a study of the residual stresses, and their postulated effects on SCC behavior, in actual canister welds are presented. Progress on the development of the model is reported.
Date: August 1, 2013
Creator: Ballinger, Ronald G.; Ferry, Sara E.; Black, Bradley P. & Teysseyre, Sebastien P.
Partner: UNT Libraries Government Documents Department

Measurement and Prediction of H2O Outgassing Kinetics from Silica-Filled Polydimethylsiloxane TR55 and S5370

Description: The isoconversional technique was employed for the measurement and prediction of H2O outgassing kinetics from silica-filled polydimethylsiloxane TR55 and S5370 in a vacuum or dry environment. Isoconversional analysis indicates that the energy barrier for H2O release from TR55 and S5370 is an increasing function of the fractional H2O release. This can be interpreted as the release of H2O from physisorbed water and then chemisorbed water with decreasing OH density from the surfaces of the embedded silica particles. Model independent predictions of H2O outgassing based on the measured kinetics follow the trend of actual isothermal outgassing at elevated temperatures, and suggest gradual outgassing in dry storage over many decades at low temperatures for both TR55 and S5370.
Date: August 28, 2006
Creator: Dinh, L N; Burnham, A K; Schildbach, M A; Smith, R A; Maxwell, R S; Balazs, B et al.
Partner: UNT Libraries Government Documents Department

Thermal Modeling of NUHOMS HSM-15 and HSM-1 Storage Modules at Calvert Cliffs Nuclear Power Station ISFSI

Description: As part of the Used Fuel Disposition Campaign of the Department of Energy (DOE), visual inspections and temperature measurements were performed on two storage modules in the Calvert Cliffs Nuclear Power Station’s Independent Spent Fuel Storage Installation (ISFSI). Detailed thermal models models were developed to obtain realistic temperature predictions for actual storage systems, in contrast to conservative and bounding design basis calculations.
Date: October 1, 2012
Creator: Suffield, Sarah R.; Fort, James A.; Adkins, Harold E.; Cuta, Judith M.; Collins, Brian A. & Siciliano, Edward R.
Partner: UNT Libraries Government Documents Department

Standard review plan for dry cask storage systems. Final report

Description: The Standard Review Plan (SRP) For Dry Cask Storage Systems provides guidance to the Nuclear Regulatory Commission staff in the Spent Fuel Project Office for performing safety reviews of dry cask storage systems. The SRP is intended to ensure the quality and uniformity of the staff reviews, present a basis for the review scope, and clarification of the regulatory requirements. Part 72, Subpart B generally specifies the information needed in a license application for the independent storage of spent nuclear fuel and high level radioactive waste. Regulatory Guide 3.61 {open_quotes}Standard Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask{close_quotes} contains an outline of the specific information required by the staff. The SRP is divided into 14 sections which reflect the standard application format. Regulatory requirements, staff positions, industry codes and standards, acceptance criteria, and other information are discussed.
Date: January 1, 1997
Partner: UNT Libraries Government Documents Department

Inspection of Used Fuel Dry Storage Casks

Description: ABSTRACT The U.S. Nuclear Regulatory Commission (NRC) regulates the storage of used nuclear fuel, which is now and will be increasingly placed in dry storage systems. Since a final disposition pathway is not defined, the fuel is expected to be maintained in dry storage well beyond the time frame originally intended. Due to knowledge gaps regarding the viability of current dry storage systems for long term use, efforts are underway to acquire the technical knowledge and tools required to understand the issues and verify the integrity of the dry storage system components. This report summarizes the initial efforts performed by researchers at Idaho National Laboratory and Argonne National Laboratory to identify and evaluate approaches to in-situ inspection dry storage casks. This task is complicated by the design of the current storage systems that severely restrict access to the casks.
Date: September 1, 2012
Creator: Kunerth, Dennis C.; McJunkin, Tim; McKay, Mark & Bakhtiari, Sasan
Partner: UNT Libraries Government Documents Department

MCO Monitoring activity description

Description: Spent Nuclear Fuel remaining from Hanford's N-Reactor operations in the 1970s has been stored under water in the K-Reactor Basins. This fuel will be repackaged, dried and stored in a new facility in the 200E Area. The safety basis for this process of retrieval, drying, and interim storage of the spent fuel has been established. The monitoring of MCOS in dry storage is a currently identified issue in the SNF Project. This plan outlines the key elements of the proposed monitoring activity. Other fuel stored in the K-Reactor Basins, including SPR fuel, will have other monitoring considerations and is not addressed by this activity description.
Date: November 9, 1998
Creator: SEXTON, R.A.
Partner: UNT Libraries Government Documents Department

Westinghouse Hanford Company special nuclear material vault storage study

Description: Category 1 and 2 Special Nuclear Materials (SNM) require storage in vault or vault type rooms as specified in DOE orders 5633.3A and 6430.1A. All category 1 and 2 SNM in dry storage on the Hanford site that is managed by Westinghouse Hanford Co (WHC) is located in the 200 West Area at Plutonium Finishing Plant (PFP) facilities. This document provides current and projected SNM vault inventories in terms of storage space filled and forecasts available space for possible future storage needs.
Date: April 23, 1996
Creator: Borisch, R.R.
Partner: UNT Libraries Government Documents Department

CPP-603 underwater fuel storage facility site integrated stabilization management plan (SISMP). Volume I

Description: The CPP-603 Underwater Fuel Storage Facility (UFSF) Site Integrated Stabilization Management Plan (SISMP) has been developed to describe the activities required for the relocation of spent nuclear fuel (SNF) from the CPP-603 facility. These activities are the only Idaho National Engineering Laboratory (INEL) actions identified in the Implementation Plan developed to meet the requirements of the Defense Nuclear Facilities Safety Board (DNFSB) remediation in the Defense Nuclear Facilities Complex. To date, 622 spent nuclear fuel units have been moved from the CPP-603 north and middle water basins, leaving 743 units in the south basin to be relocated from the facility by December 31, 2000. Besides moving fuels from the CPP-603, in 1993 and 1994 more than 300 fuel storage yokes in the north and middle basins were redundantly rigged because of corrosion problems. More than 200 fuel transfers within the north and middle basins were also made to ensure proper spacing of the fuels, and 104 corroded cans containing spent space reactor fuel were repackaged underwater to prevent potential release of their contents. This document is provided to address the relocation activities for the remaining 743 units in the south basin into wet storage pools at building CPP-666 or into dry storage at the Irradiation Fuel Storage Facility (IFSF).
Date: September 1, 1996
Creator: Wachs, G.W.; Blake, H.M.; Cottam, R.E.; Denney, R.D. & Shiffern, R.A.
Partner: UNT Libraries Government Documents Department

Heat Transfer Modeling of Dry Spent Nuclear Fuel Storage Facilities

Description: The present work was undertaken to provide heat transfer model that accurately predicts the thermal performance of dry spent nuclear fuel storage facilities. One of the storage configurations being considered for DOE Aluminum-clad Spent Nuclear Fuel (Al-SNF), such as the Material and Testing Reactor (MTR) fuel, is in a dry storage facility. To support design studies of storage options a computational and experimental program has been conducted at the Savannah River Site (SRS). The main objective is to develop heat transfer models including natural convection effects internal to an interim dry storage canister and to geological codisposal Waste Package (WP). Calculated temperatures will be used to demonstrate engineering viability of a dry storage option in enclosed interim storage and geological repository WP and to assess the chemical and physical behaviors of the Al-SNF in the dry storage facilities. The current paper describes the modeling approaches and presents the computational results along with the experimental data.
Date: January 13, 1999
Creator: Lee, S.Y.
Partner: UNT Libraries Government Documents Department

Characterization program management plan for Hanford K Basin Spent Nuclear Fuel

Description: A management plan was developed for Westinghouse Hanford Company (WHC) and Pacific Northwest Laboratories (PNL) to work together on a program to provide characterization data to support removal, conditioning and subsequent dry storage of the spent nuclear fuels stored at the Hanford K Basins. The Program initially supports gathering data to establish the current state of the fuel in the two basins. Data Collected during this initial effort will apply to all SNF Project objectives. N Reactor fuel has been degrading with extended storage resulting in release of material to the basin water in K East and to the closed conisters in K West. Characterization of the condition of these materials and their responses to various conditioning processes and dry storage environments are necessary to support disposition decisions. Characterization will utilize the expertise and capabilities of WHC and PNL organizations to support the Spent Nuclear Fuels Project goals and objectives. This Management Plan defines the structure and establishes the roles for the participants providing the framework for WHC and PNL to support the Spent Nuclear Fuels Project at Hanford
Date: October 18, 1995
Creator: Lawrence, L.A.
Partner: UNT Libraries Government Documents Department

Natural convection heat transfer within horizontal spent nuclear fuel assemblies

Description: Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.
Date: December 1, 1995
Creator: Canaan, R.E.
Partner: UNT Libraries Government Documents Department

Radiation doses in granite around emplacement holes in the Spent Fuel Test - Climax. Final report

Description: Final comparisons are made between measured and calculated radiation doses around the holes in which the spent fuel was emplaced in the Spent Fuel Test - Climax. Neutron doses were found to be negligible compared with gamma doses. Good agreement was found between the doses predicted by Monte Carlo calculations and those measured by short-exposure thermoluminescence dosimetry. Poor agreement was found between the calculational results and doses measured by exposure of LiF optical-absorption-type dosimeters for long periods, probably because of an inability to accurately correct for fade resulting from elevated temperature exposure over several months. The maximum dose to the rock occurred at the walls of the emplacement holes, and amounted to 1.6 MGy (1.6 x 10{sup 8} rad) in granite for the emplacement period of nearly 3 years. It is recommended that dose evaluations for future high-level nuclear waste storage facilities also be performed by combining calculations and dosimetry. Passive dosimetry techniques, if used, should involve short exposures, so that laboratory calibrations can be performed with duplicate time, temperature, dose rate, and dose parameters. An attractive alternative would be to use active ionization chambers, inserted only periodically. These could be calibrated under appropriate temperature and pressure conditions, and could be read directly. 23 references, 7 figures, 8 tables.
Date: July 26, 1984
Creator: Van Konynenburg, R.A.
Partner: UNT Libraries Government Documents Department

Conceptual design report for the ICPP spent nuclear fuel dry storage project

Description: The conceptual design is presented for a facility to transfer spent nuclear fuel from shipping casks to dry storage containers, and to safely store those containers at ICPP at INEL. The spent fuels to be handled at the new facility are identified and overall design and operating criteria established. Physical configuration of the facility and the systems used to handle the SNF are described. Detailed cost estimate for design and construction of the facility is presented.
Date: July 1, 1996
Partner: UNT Libraries Government Documents Department

Technical basis for storage of Zircaloy-clad spent fuel in inert gases

Description: This report summarizes the technical bases to establish safe conditions for dry storage of Zircaloy-clad fuel. Dry storage of fuel with zirconium alloy cladding has been licensed in Canada, the Federal Republic of Germany, and Switzerland. In addition, dry storage demonstrations, hot cell tests, and modeling have been conducted using Zircaloy-clad fuel. The demonstrations have included irradiated boiling water reactor, pressurized heavy-water reactor, and pressurized water reactor (PWR) fuel assemblies. Irradiated fuel has been emplaced in and retrieved from metal casks, dry wells, silos, and a vault. Dry storage tests and demonstrations have involved {similar_to}5,000 fuel rods, and {similar_to}600 rods have been monitored during dry storage in inert gases with maximum cladding temperatures ranging from 50 to 570{sup 0}C. Although some tests and demonstrations are still in progress, there is currently no evidence that any rods exposed to inert gases have failed (one PWR rod exposed to an air cover gas failed at {similar_to}70{sup 0}C). Based on this favorable experience, it is concluded that there is sufficient information on fuel rod behavior, storage conditions, and potential cladding failure mechanisms to support licensing of dry storage in the United States. This licensing position includes a requirement for inert cover gases and a maximum cladding temperature guideline of 380{sup 0}C for Zircaloy-clad fuel. Using an inert cover gas assures that even if fuel with cladding defects were placed in dry storage, or if defects develop during storage, the defects would not propagate. Tests and demonstrations involving Zircaloy-clad rods and assemblies with maximum cladding temperatures above 400{sup 0}C are in progress. When the results from these tests have been evaluated, the viability of higher temperature limits should be examined. Acceptable conditions for storage in air and dry storage of consolidated fuel are issues yet to be resolved.
Date: September 1, 1983
Creator: Johnson, A.B. Jr. & Gilbert, E.R.
Partner: UNT Libraries Government Documents Department

Technical issues for possible dry storage of DOE owned spent nuclear fuels

Description: Criteria for interim dry storage of DOE owned spent nuclear fuels will be based on a combination of technical, regulatory, and political requirements. These requirements have not been fully established but will include issues such as the necessity for monitoring the fuel, retrievability of the fuel, maintenance of final disposal options, siting of interim disposal facilities (will the fuels at each site remain on that site throughout the interim disposal period?), and interfacing with the International Atomic Energy Agency over inspectability of the highly enriched fuels in the inventory. Regardless of the other requirement issues, the technical storage criteria are likely to vary with fuel type and will include: (a) a fuel specific maximum storage temperature coupled to; (b) a specific storage environment (for example, Helium gas with less than xx ppm moisture); (c) a mass limit for fissionable isotopes; and (d) provisions for prestorage characterization and, if necessary, segregation of the fuel based on the established condition of the fuel. Each of these criteria has several associated technical issues and there are significant interactions among the criteria. This white paper summarizes many of the technical issues which must be resolved in order to develop functional and design specification for the interim dry storage for DOE owned spent nuclear fuels.
Date: July 1, 1995
Creator: McKinnon, M.A.; Einziger, R.E.; Louthan, M.L. & Iyer, N.C.
Partner: UNT Libraries Government Documents Department

Preliminary Design Report Shippingport Spent Fuel Drying and Inerting System

Description: A process description and system flow sheets have been prepared to support the design/build package for the Shippingport Spent Fuel Canister drying and inerting process skid. A process flow diagram was prepared to show the general steps to dry and inert the Shippingport fuel loaded into SSFCs for transport and dry storage. Flow sheets have been prepared to show the flows and conditions for the various steps of the drying and inerting process. Calculations and data supporting the development of the flow sheets are included.
Date: May 18, 2000
Creator: JEPPSON, D.W.
Partner: UNT Libraries Government Documents Department

Drying results of K-Basin fuel element 0309M (Run 3)

Description: An N-Reactor outer fuel element that had been stored underwater in the Hanford 100 Area K-West Basin was subjected to a combination of low- and high-temperature vacuum drying treatments. These studies are part of a series of tests being conducted by Pacific Northwest National Laboratory on the drying behavior of spent nuclear fuel elements removed from both the K-West and K-East Basins. The drying test series was designed to test fuel elements that ranged from intact to severely damaged. The fuel element discussed in this report was removed from K-West canister 0309M during the second fuel selection campaign, conducted in 1996, and has remained in wet storage in the Postirradiation Testing Laboratory (PTL, 327 Building) since that time. The fuel element was broken in two pieces, with a relatively clean fracture, and the larger piece was tested. A gray/white coating was observed. This was the first test of a damaged fuel element in the furnace. K-West canisters can hold up to seven complete fuel assemblies, but, for purposes of this report, the element tested here is designated as Element 0309M. Element 0309M was subjected to drying processes based on those proposed under the Integrated Process Strategy, which included a hot drying step.
Date: July 1, 1998
Creator: Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J. & Ritter, G.A.
Partner: UNT Libraries Government Documents Department

System design description for the whole element furnace testing system

Description: This document provides a detailed description of the Hanford Spent Nuclear Fuel (SNF) Whole Element Furnace Testing System located in the Postirradiation Testing Laboratory G-Cell (327 Building). Equipment specifications, system schematics, general operating modes, maintenance and calibration requirements, and other supporting information are provided in this document. This system was developed for performing cold vacuum drying and hot vacuum drying testing of whole N-Reactor fuel elements, which were sampled from the 105-K East and K West Basins. The proposed drying processes are intended to allow dry storage of the SNF for long periods of time. The furnace testing system is used to evaluate these processes by simulating drying sequences with a single fuel element and measuring key system parameters such as internal pressures, temperatures, moisture levels, and off-gas composition.
Date: May 1, 1998
Creator: Ritter, G.A.; Marschman, S.C.; MacFarlan, P.J. & King, D.A.
Partner: UNT Libraries Government Documents Department

Analysis of Dry Storage Temperature Limits for Zircaloy-Clad Spent Nuclear Fuel

Description: Safe interim dry storage of spent nuclear fuel (SNF) must be maintained for a minimum of twenty years according to the Code of Federal Regulations. The most important variable that must be regulated by dry storage licensees in order to meet current safety standards is the temperature of the SNF. The two currently accepted models to define the maximum allowable initial storage temperature for SNF are based on the diffusion controlled cavity growth (DCCG) failure mechanism proposed by Raj and Ashby. These models may not give conservative temperature limits. Some have suggested using a strain-based failure model to predict the maximum allowable temperatures, but we have shown that this is not applicable to the SNF as long as DCCG is the assumed failure mechanism. Although the two accepted models are based on the same fundamental failure theory (DCCG), the researchers who developed the models made different assumptions, including selection of some of the most critical variables in the DCCG failure equation. These inconsistencies are discussed together with recommended modifications to the failure models based on more recent data.
Date: November 5, 1999
Creator: Hayes, T.A.; Kassner, M.D. & Vecchio, K.S.
Partner: UNT Libraries Government Documents Department

Technical Issues and Characterization for Fuel and Sludge in Hanford K Basins

Description: Technical Issues for the interim dry storage of N Reactor Spent Nuclear Fuel (SNF) are discussed. Characterization data from fuel, to support resolution of these issues, are reviewed and new results for the oxidation of fuel in a moist atmosphere and the drying of whole fuel elements are presented. Characterization of associated K basin sludge is also discussed in light of a newly adopted disposal pathway.
Date: June 1, 2000
Creator: MAKENAS, B.J.
Partner: UNT Libraries Government Documents Department