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Method of estimating maximum VOC concentration in void volume of vented waste drums using limited sampling data: Application in transuranic waste drums

Description: A test program has been conducted at the Idaho National Engineering Laboratory to demonstrate that the concentration of volatile organic compounds (VOCs) within the innermost layer of confinement in a vented waste drum can be estimated using a model incorporating diffusion and permeation transport principles as well as limited waste drum sampling data. The model consists of a series of material balance equations describing steady-state VOC transport from each distinct void volume in the drum. The primary model input is the measured drum headspace VOC concentration. Model parameters are determined or estimated based on available process knowledge. The model effectiveness in estimating VOC concentration in the headspace of the innermost layer of confinement was examined for vented waste drums containing different waste types and configurations. This paper summarizes the experimental measurements and model predictions in vented transuranic waste drums containing solidified sludges and solid waste.
Date: December 1, 1995
Creator: Liekhus, K.J. & Connolly, M.J.
Partner: UNT Libraries Government Documents Department

Idaho National Engineering Laboratory radiological control performance indicator report. Fourth quarterly calendar year 1994

Description: This document provides a report and analysis of the Radiological Control Program through the fourth quarter of calendar year 1994 (CY-1994) at the Idaho National Engineering Laboratory (INEL) under the direction of Lockheed Idaho Technologies Company (LITCO). The Radiological Performance Indicator Report is provided in accordance with Article 133 of the INEL Radiological Control Manual.
Date: April 1, 1995
Creator: Aitken, S.B.
Partner: UNT Libraries Government Documents Department

Macroencapsulation of lead and steel SWARF

Description: The treatability study to macroencapsulate radioactively contaminated lead and steel swarf (cuttings and/or chips)and chunks, a low level mixed waste, from the dismantlement of excess surplus uranium fuel handling and transfer casks was successful. Macroencapsulation is the land disposal restriction treatment standard for this waste form per 40 CFR 268.42 Table 3. An epoxy-based thermoset system was employed due to cracking failures of other types of thermoset systems. Bench scale tests were performed with a two-part epoxy (resin and hardener) using cast iron chips as a surrogate waste media. A two stage encapsulation process was employed in treating the swarf. Two liters of epoxy were added to a 2.8{ell} (3 qt) container of swarf under 51K Pa vacuum (-15-inch of Hg) during the first stage of the process. In this stage each individual particle or chip was wetted by epoxy and allowed to harden into an initial monolith. The second stage encapsulated the initial monolith with a secondary layer of epoxy forming a larger final monolith. By evacuating the air from the swarf and epoxy during the initial monolith encapsulation, a higher density (higher swarf to epoxy ratio) was achieved. Tensile and compressive strength tests were performed on samples and without any media (cast iron chips). The coupons were prepared from a series of monoliths featuring various mixtures ratios and vacuum levels. The tensile strength of epoxy without chips averaged 41M Pa (6000 psi) and 1.4M Pa (2000 psi) with cast iron chips. Compression strengths averaged 140M Pa (20,000 psi) without chips and 66.2M Pa (9600 psi) with cast iron chips.
Date: December 1, 1995
Creator: Zirker, L.; Thiesen, T.; Tyson, D. & Beitel, G.
Partner: UNT Libraries Government Documents Department

National Low-Level Waste Management Program Radionuclide Report Series: Volume 12, Cobalt-60

Description: This report outlines the basic radiological and chemical characteristics of cobalt-60 ({sup 60}Co) and examines how these characteristics affect the behavior of {sup 60}Co in various environmental media, such as soils, groundwater, plants, animals, the atmosphere, and the human body. Discussions also include methods of {sup 60}Co production, waste types, and waste forms that contain {sup 60}Co. All cobalt atoms contain 27 protons (Z = 27) and various numbers of neutrons (typically N = 27 to 37 neutrons) within the atom`s nucleus. There is only one stable isotope of cobalt, namely {sup 59}Co. All other cobalt isotopes, including {sup 60}Co, are radioactive. The radioactive isotopes of cobalt have half-lives ranging from less than a second ({sup 54}Co-0.19 s) to 5.2 years ({sup 60}Co). The radioactive isotopes of cobalt are not a normal constituent of the natural environment and are generated as a result of human activities. The primary source of {sup 60}Co in the environment has been low-level radioactive waste material generated as a result of neutron activation of stable {sup 59}Co that is present in the structural components of nuclear reactor vessels. This isotope is also intentionally produced, usually in reactors but also to some degree in accelerators for industrial and medical uses, such as for radiation sources for cancer treatment and nondestructive testing of metals and welds. {sup 60}Co may enter the environment as a result of the activities associated with nuclear reactor operations and decommissioning and when industrial and medical sources are being used, manufactured, or disposed.
Date: June 1, 1995
Creator: Adams, J.P.
Partner: UNT Libraries Government Documents Department

TMI-2 analysis using SCDAP/RELAP5/MOD3.1

Description: SCDAP/RELAP5/MOD3.1, an integrated thermal hydraulic analysis code developed primarily to simulate severe accidents in nuclear power plants, was used to predict the progression of core damage during the TMI-2 accident. The version of the code used for the TMI-2 analysis described in this paper includes models to predict core heatup, core geometry changes, and the relocation of molten core debris to the lower plenum of the reactor vessel. This paper describes the TMI-2 input model, initial conditions, boundary conditions, and the results from the best-estimate simulation of Phases 1 to 4 of the TMI-2 accident as well as the results from several sensitivity calculations.
Date: November 1, 1994
Creator: Hohorst, J.K.; Polkinghorne, S.T.; Siefken, L.J.; Allison, C.M. & Dobbe, C.A.
Partner: UNT Libraries Government Documents Department

Fusion Safety Program annual report, fiscal year 1994

Description: This report summarizes the major activities of the Fusion Safety Program in fiscal year 1994. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory and Lockheed Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in 1979. Activities are conducted at the INEL, at other DOE laboratories, and at other institutions, including the University of Wisconsin. The technical areas covered in this report include tritium safety, beryllium safety, chemical reactions and activation product release, safety aspects of fusion magnet systems, plasma disruptions, risk assessment failure rate data base development, and thermalhydraulics code development and their application to fusion safety issues. Much of this work has been done in support of the International Thermonuclear Experimental Reactor (ITER). Also included in the report are summaries of the safety and environmental studies performed by the Fusion Safety Program for the Tokamak Physics Experiment and the Tokamak Fusion Test Reactor and of the technical support for commercial fusion facility conceptual design studies. A major activity this year has been work to develop a DOE Technical Standard for the safety of fusion test facilities.
Date: March 1, 1995
Creator: Longhurst, G.R.; Cadwallader, L.C.; Dolan, T.J.; Herring, J.S.; McCarthy, K.A.; Merrill, B.J. et al.
Partner: UNT Libraries Government Documents Department

Remediating the INEL`s buried mixed waste tanks

Description: The Idaho National Engineering Laboratory (INEL), formerly the National Reactor Testing Station (NRTS), encompasses 890 square miles and is located in southeast Idaho. In 1949, the United States Atomic Energy Commission, now the Department of Energy (DOE), established the NRTS as a site for the building and testing of nuclear facilities. Wastes generated during the building and testing of these nuclear facilities were disposed within the boundaries of the site. These mixed wastes, containing radionuclides and hazardous materials, were often stored in underground tanks for future disposal. The INEL has 11 buried mixed waste storage tanks regulated under the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) ranging in size from 400 to 50,000 gallons. These tanks are constructed of either stainless or carbon steel and are located at 3 distinct geographic locations across the INEL. These tanks have been grouped based on their similarities in an effort to save money and decrease the time required to complete the necessary remediation. Environmental Restoration and Technology Development personnel are teaming in an effort to address the remediation problem systematically.
Date: February 28, 1996
Creator: Kuhns, D.J.; Matthern, G.E. & Reese, C.L.
Partner: UNT Libraries Government Documents Department

US hydropower resource assessment for Wisconsin

Description: The Department of Energy is developing an estimate of the undeveloped hydropower potential in this country. The Hydropower Evaluation Software is a computer model that was developed by the Idaho National Engineering Laboratory for this purpose. The software measures the undeveloped hydropower resources available in the United States, using uniform criteria for measurement. The software was developed and tested using hydropower information and data provided by the Southwestern Power Administration. It is a menu-driven software program that allows the personal computer user to assign environmental attributes to potential hydropower sites, calculate development suitability factors for each site based on the environmental attributes present, and generate reports based on these suitability factors. This report details the resource assessment results for the State of Wisconsin.
Date: May 1, 1996
Creator: Conner, A.M. & Francfort, J.E.
Partner: UNT Libraries Government Documents Department

Decomposition of PCBs in oils using gamma radiolysis: A treatability study. Final report

Description: This report presents the results of a treatability study of radiologically and PCB contaminated waste hydraulic oils at the Idaho National Engineering Laboratory (INEL). The goal of the study was to demonstrate that PCBs could be selectively removed from the contaminated oils. The PCBs were selectively decomposed in an in-situ fashion via gamma-ray radiolysis. The gamma-ray source was spent nuclear fuel at the Advanced Test Reactor (ATR) canal at the Test Reactor Area (TRA), of the INEL. Exposure to gamma-rays does not induce radioactivity in the exposed solutions. The treatability study was the culmination of five years of research concerning PCB radiolysis conducted at INEL which investigated the mechanism and kinetics of the reaction in several solvents. The major findings of this research are summarized here. Based upon these findings three INEL waste streams were selected for testing of the process. The Environmental Protection Agency (EPA) treatment standard of 2 mg/kg was successfully achieved in all waste streams. The interference of contaminants other than PCBs is discussed.
Date: April 1, 1996
Creator: Mincher, B.J. & Arbon, R.E.
Partner: UNT Libraries Government Documents Department

Inverse modeling for field-scale hydrologic and transport parameters of fractured basalt

Description: A large-scale test of infiltration into a thick sequence Of fractured Snake River Plain basalts was performed during the summer of 1994 on the Idaho National Engineering Laboratory. Monitoring of moisture and tracer movement during this test provided a set of quantitative measurements from which to obtain a field-scale hydrologic description of the fractured basalts. An inverse modeling study using these quantitative measurements was performed to obtain the representative hydrologic description. This report describes the results of the inverse modeling study and includes the background and motivation for conducting the infiltration test; a brief overview of the infiltration test; descriptions of the calibration targets chosen for the simulation study, the simulation model, and the model implementation; and the simulation results with comparisons to hydrologic and tracer breakthrough data obtained from the infiltration test.
Date: December 1, 1995
Creator: Magnuson, S.O.
Partner: UNT Libraries Government Documents Department

Measurement of {sup 222}Rn flux, {sup 222}Rn emanation and {sup 226}Ra concentration from injection well pipe scale

Description: The presence of Naturally Occurring Radioactive Material (NORM) has been recognized since the early 1930s in petroleum reservoirs and in oil and gas production and processing facilities. NORM was typically observed in barite scale that accumulated on the interior of oil production tubing and in storage tank and heater-treater separation sludge. Recent concern has been expressed over the health impacts from the uncontrolled release of NORM to the public. There are several potential exposure pathways to humans from oil-field NORM. Among these is inhalation of radon gas and its daughter products. For this exposure pathway to be of any significance, radon must first be released from the NORM matrix and diffuse in free air. The radon emanation fraction refers to the fraction of radon atoms produced by the decay of radium, that migrate from the bulk material as free gaseous atoms. The purpose of this investigation was to characterize the radon release rates from NORM-scale contaminated production tubing being stored above ground, characterize the radon emanation fraction of the bulk scale material when removed from the tubing, and characterize the radium concentrations of the scale. Accurate characterization of {sup 222}Rn emanation fractions from pipe scale may dictate the type of disposal options available for this waste. Characterization of radon release from stored pipes will assist in determining if controls are needed for workers or members of the public downwind from the source. Due to the sensitive nature of this data, the location of this facility is not disclosed.
Date: February 1996
Creator: Rood, A. S. & Kendrick, D. T.
Partner: UNT Libraries Government Documents Department

Overview of groundwater and surface water standards pertinent to the Idaho National Engineering Laboratory. Revision 3

Description: This document presents an overview of groundwater- and surface water-related laws, regulations, agreements, guidance documents, Executive Orders, and DOE orders pertinent to the Idaho National Engineering Laboratory. This document is a summary and is intended to help readers understand which regulatory requirements may apply to their particular circumstances. However, the document is not intended to be used in lieu of applicable regulations. Unless otherwise noted, the information in this report reflects a summary and evaluation completed July 1, 1995. This document is considered a Living Document, and updates on changing laws and regulations will be provided.
Date: September 1, 1995
Creator: Lundahl, A.L.; Williams, S. & Grizzle, B.J.
Partner: UNT Libraries Government Documents Department

Summary of treatment, storage, and disposal facility usage data collected from U.S. Department of Energy sites

Description: This report presents an analysis for the US Department of Energy (DOE) to determine the level and extent of treatment, storage, and disposal facility (TSDF) assessment duplication. Commercial TSDFs are used as an integral part of the hazardous waste management process for those DOE sites that generate hazardous waste. Data regarding the DOE sites` usage have been extracted from three sets of data and analyzed in this report. The data are presented both qualitatively and quantitatively, as appropriate. This information provides the basis for further analysis of assessment duplication to be documented in issue papers as appropriate. Once the issues have been identified and adequately defined, corrective measures will be proposed and subsequently implemented.
Date: April 1, 1995
Creator: Jacobs, A.; Oswald, K. & Trump, C.
Partner: UNT Libraries Government Documents Department

Development and testing of a SREX flowsheet for the partitioning of strontium and lead from simulated ICPP sodium-bearing waste

Description: Laboratory experimentation has indicated that the SREX process is effective for partitioning {sup 90}Sr from acidic radioactive waste solutions located at the Idaho Chemical Processing Plant. Previous countercurrent flowsheet testing of the SREX process with simulated waste resulted in 99.98% removal of Sr. With this previous test, however, Pb was extracted by the SREX solvent and was not back-extracted in the dilute nitric acid strip section. The Pb concentration increased in the recycled solvent and in the aqueous phase of the strip section, resulting in the formation of a Pb precipitate. Subsequently, studies were initiated to identify alternative stripping agents which will selectively strip Sr and Pb from the SREX solvent. Based on the results of these studies, a countercurrent flow sheet was developed and tested in the 5.5-cm Centrifugal Contactor Mockup using simulated waste. The flowsheet tested consisted of an extraction section (0.15 M 4{prime},4{prime}(5)-di-(tert-butyldicyclohexo)-18-crown-6 and 1.2 M TBP in Isopar-L{reg_sign}), a 0.05 M nitric acid strip section for the removal of Sr from the SREX solvent, a 0.1 M ammonium citrate strip section for the removal of Pb from the SREX solvent, and a 2.0 M nitric acid equilibration section. The behavior of Sr, Pb, Al, Ca, Hg, Na, Zr, and H{sup +} was evaluated. The described flowsheet successfully extracted and selectively stripped Sr and Pb from the SBW simulant. Removal efficiencies of 97.9% and 99.91% were obtained for Sr and Pb, respectively. Essentially all of the extracted Sr (99.998%) and 1.9% of extracted Pb exited with the 0.05 M nitric acid strip product; whereas, 0.002% of the extracted Sr and 97.9% of the extracted Pb existed with the 0.1 M ammonium citrate strip product. Also, 95% of the Hg and 63% of the Zr were extracted by the SREX solvent.
Date: November 1, 1996
Creator: Law, J.D. & Wood, D.J.
Partner: UNT Libraries Government Documents Department

Modeling and design of a reload PWR core for a 48-month fuel cycle

Description: The objective of this research was to use state-of-the-art nuclear and fuel performance packages to evaluate the feasibility and costs of a 48 calendar month core in existing pressurized water reactor (PWR) designs, considering the full range of practical design and economic considerations. The driving force behind this research is the desire to make nuclear power more economically competitive with fossil fuel options by expanding the scope for achievement of higher capacity factors. Using CASMO/SIMULATE, a core design with fuel enriched to 7{sup w}/{sub o} U{sup 235} for a single batch loaded, 48-month fuel cycle has been developed. This core achieves an ultra-long cycle length without exceeding current fuel burnup limits. The design uses two different types of burnable poisons. Gadolinium in the form of gadolinium oxide (Gd{sub 2}O{sub 3}) mixed with the UO{sub 2} of selected pins is sued to hold down initial reactivity and to control flux peaking throughout the life of the core. A zirconium di-boride (ZrB{sub 2}) integral fuel burnable absorber (IFBA) coating on the Gd{sub 2}O{sub 3}-UO{sub 2} fuel pellets is added to reduce the critical soluble boron concentration in the reactor coolant to within acceptable limits. Fuel performance issues of concern to this design are also outlined and areas which will require further research are highlighted.
Date: May 1997
Creator: McMahon, M. V.; Driscoll, M. J. & Todreas, N. E.
Partner: UNT Libraries Government Documents Department

Evaluation of tubular reactor designs for supercritical water oxidation of U.S. Department of Energy mixed waste

Description: Supercritical water oxidation (SCWO) is an emerging technology for industrial waste treatment and is being developed for treatment of the US Department of Energy (DOE) mixed hazardous and radioactive wastes. In the SCWO process, wastes containing organic material are oxidized in the presence of water at conditions of temperature and pressure above the critical point of water, 374 C and 22.1 MPa. DOE mixed wastes consist of a broad spectrum of liquids, sludges, and solids containing a wide variety of organic components plus inorganic components including radionuclides. This report is a review and evaluation of tubular reactor designs for supercritical water oxidation of US Department of Energy mixed waste. Tubular reactors are evaluated against requirements for treatment of US Department of Energy mixed waste. Requirements that play major roles in the evaluation include achieving acceptable corrosion, deposition, and heat removal rates. A general evaluation is made of tubular reactors and specific reactors are discussed. Based on the evaluations, recommendations are made regarding continued development of supercritical water oxidation reactors for US Department of Energy mixed waste.
Date: December 1994
Creator: Barnes, C. M.
Partner: UNT Libraries Government Documents Department

PST user`s guide

Description: The Parametric Source Term (PST) software allows estimation of radioactivity release fractions for Level 2 Probabilistic Safety Assessments (PSAs). PST was developed at the Idaho National Engineering Laboratory (INEL) for the Nuclear Regulatory Commission`s (NRC`s) Accident Sequence Precursor (ASP) Program. PST contains a framework of equations that model activity transport between volumes in the release pathway from the core, through the vessel, through the containment, and to the environment. PST quickly obtains exact solutions to differential equations for activity transport in each volume for each time interval. PST provides a superior method for source term estimation because it: ensures conservation of activity transported across various volumes in the release pathway; provides limited consideration of the time-dependent behavior of input parameter uncertainty distributions; allows input to be quantified using state-of-the-art severe accident analysis code results; increases modeling flexibility because linkage between volumes is specified by user input; and allows other types of Light Water Reactor (LWR) plant designs to be evaluated with minimal modifications. PST is a microcomputer-based system that allows the analyst more flexibility than a mainframe system. PST has been developed to run with both MS DOS and MS Windows 95/NT operating systems. PST has the capability to load ASP Source Term Vector (STV) information, import pre-specified default input for the 6 Pressurized Water Reactors (PWRs) initially analyzed in the NRC ASP program, allow input value modifications for release fraction sensitivity studies, export user-specified default input for the LWR being modeled, report results of radioactivity release calculations at each time interval, and generate formatted results that can interface with other risk assessment codes. This report describes the PST model and provides guidelines for using PST.
Date: October 1, 1996
Creator: Rempe, J.L.; Cebull, M.J. & Gilbert, B.G.
Partner: UNT Libraries Government Documents Department

The potential pyrophoricity of BMI-SPEC and aluminum plate spent fuels retrieved from underwater storage

Description: Physical/chemical factors in U metal and hydride combustion, particularly pyrophoricity in ambient environment, were evaluated for BMI-SPEC and UAl{sub x} plate fuels. Some metal fuels may be highly reactive (spontaneously igniting in air) due to high specific surface area, high decay heat, or a high U hydride content from corrosion during underwater storage. However, for the BMI-SPEC and the aluminum plate fuels, this reactivity is too low to present a realistic threat of uncontrolled spontaneous combustion at ambient conditions. While residual U hydride is expected in these corroded fuels, the hydride levels are expected to be too low and the configuration too unfavorable to ignite the fuel meat when the fuels are retrieved from the basin and dried. Furthermore the composition and microstructure of the UAl{sub x} fuels further mitigate that risk.
Date: August 1996
Creator: Ebner, M. A.
Partner: UNT Libraries Government Documents Department

Avoidable waste management costs

Description: This report describes the activity based costing method used to acquire variable (volume dependent or avoidable) waste management cost data for routine operations at Department of Energy (DOE) facilities. Waste volumes from environmental restoration, facility stabilization activities, and legacy waste were specifically excluded from this effort. A core team consisting of Idaho National Engineering Laboratory, Los Alamos National Laboratory, Rocky Flats Environmental Technology Site, and Oak Ridge Reservation developed and piloted the methodology, which can be used to determine avoidable waste management costs. The method developed to gather information was based on activity based costing, which is a common industrial engineering technique. Sites submitted separate flow diagrams that showed the progression of work from activity to activity for each waste type or treatability group. Each activity on a flow diagram was described in a narrative, which detailed the scope of the activity. Labor and material costs based on a unit quantity of waste being processed were then summed to generate a total cost for that flow diagram. Cross-complex values were calculated by determining a weighted average for each waste type or treatability group based on the volume generated. This study will provide DOE and contractors with a better understanding of waste management processes and their associated costs. Other potential benefits include providing cost data for sites to perform consistent cost/benefit analysis of waste minimization and pollution prevention (WMIN/PP) options identified during pollution prevention opportunity assessments and providing a means for prioritizing and allocating limited resources for WMIN/PP.
Date: January 1, 1995
Creator: Hsu, K.; Burns, M.; Priebe, S. & Robinson, P.
Partner: UNT Libraries Government Documents Department

Chemical and mechanical performance properties for various final waste forms -- PSPI scoping study

Description: The US DOE is obtaining data on the performance properties of the various final waste forms that may be chosen as primary treatment products for the alpha-contaminated low-level and transuranic waste at the INEL`s Transuranic Storage Area. This report collects and compares selected properties that are key indicators of mechanical and chemical durability for Portland cement concrete, concrete formed under elevated temperature and pressure, sulfur polymer cement, borosilicate glass, and various forms of alumino-silicate glass, including in situ vitrification glass and various compositions of iron-enriched basalt (IEB) and iron-enriched basalt IV (IEB4). Compressive strength and impact resistance properties were used as performance indicators in comparative evaluation of the mechanical durability of each waste form, while various leachability data were used in comparative evaluation of each waste form`s chemical durability. The vitrified waste forms were generally more durable than the non-vitrified waste forms, with the iron-enriched alumino-silicate glasses and glass/ceramics exhibiting the most favorable chemical and mechanical durabilities. It appears that the addition of zirconia and titania to IEB (forming IEB4) increases the leach resistance of the lanthanides. The large compositional ranges for IEB and IEB4 more easily accommodate the compositions of the waste stored at the INEL than does the composition of borosilicate glass. It appears, however, that the large potential variation in IEB and IEB4 compositions resulting from differing waste feed compositions can impact waste form durability. Further work is needed to determine the range of waste stream feed compositions and rates of waste form cooling that will result in acceptable and optimized IEB or IEB4 waste form performance. 43 refs.
Date: September 1996
Creator: Farnsworth, R. K.; Larsen, E. D.; Sears, J. W.; Eddy, T. L. & Anderson, G. L.
Partner: UNT Libraries Government Documents Department

Cooperative Telerobotic Retrieval system Phase 1 technology evaluation report

Description: This document describes the results from the Cooperative Telerobotic Retrieval demonstration and testing conducted at the Idaho National Engineering Laboratory during December 1994 and January 1995. The purpose of the demonstration was to ascertain the feasibility of the system for deploying tools both independently and cooperatively for supporting remote characterization and removal of buried waste in a safe manner and in compliance with all regulatory requirements. The procedures and goals of the demonstration were previously defined in the Cooperative Telerobotic Retrieval System Test Plan for Fiscal Year 1994, which served as a guideline for evaluating the system.
Date: March 1, 1995
Creator: Hyde, R.A. & Croft, K.M.
Partner: UNT Libraries Government Documents Department

Environment, safety, health, and quality plan for the TRU- Contaminated Arid Soils Project of the Landfill Stabilization Focus Area Program

Description: The Landfill Stabilization Focus Area (LSFA) is a program funded by the US Department of Energy Office of Technology Development. LSFA supports the applied research, development, demonstration, testing, and evaluation of a suite of advanced technologies that together form a comprehensive remediation system for the effective and efficient remediation of buried waste. The TRU-Contaminated Arid Soils project is being conducted under the auspices of the LSFA Program. This document describes the Environment, Safety, Health, and Quality requirements for conducting LSFA/Arid Soils activities at the Idaho National Engineering Laboratory. Topics discussed in this report, as they apply to LSFA/Arid Soils operations, include Federal, State of Idaho, and Environmental Protection Agency regulations, Health and Safety Plans, Quality Program, Data Quality Objectives, and training and job hazard analysis. Finally, a discussion is given on CERCLA criteria and system and performance audits as they apply to the LSFA Program.
Date: June 1, 1995
Creator: Watson, L.R.
Partner: UNT Libraries Government Documents Department

Slurry-based fabrication of chopped fiberglass composite preforms

Description: A water-based process for the fabrication of chopped fiberglass preforms is being developed in collaboration with the Automotive Composite Consortium (ACC) and The Budd Company. This slurry process uses hydraulic pressure to form highly compacted fiberglass preforms on contoured, perforated metal screens. The preforms will be used in the development of structural automotive composites. A key objective is to produce preforms having uniform areal density. Computational simulation of variable open area screens, and areal density mapping using a gamma densitometer are discussed.
Date: December 1, 1995
Creator: Moore, G.A.; Johnson, R.W.; Landon, M.D.; Stoots, C.M. & Anderson, J.L.
Partner: UNT Libraries Government Documents Department

Radioactive liquid waste generation goals at the ICPP

Description: Processes at ICPP generating hazardous radioactive liquid wastes (which are sent to the tank farm) include NWCF, PEW evaporator, LET&D, tank farm, fuel storage operations, etc. In May 1994, the INEL Radioactive Liquid Waste Management Plan was issued but not implemented. Waste generation goals have been reviewed and updated in this report (details are given in appendix). A meeting was held to determine the new waste generation goals and best approach to reaching them. Waste streams were individually analyzed in this meeting and several adjustments made both during the meeting and following the meeting. The information was adjusted and modeling completed to determine the waste reduction achieved (spreadsheets are included in appendix). Results of this update indicate that there has been a significant reduction in the waste generation goals from 2 years ago. If the updated baseline goals are met, a 35% waste reduction will be achieved; this coupled with increased calcination rate, will enable the waste in the tank farm to be processed by 2012; however a program is needed to ensure these waste goals are met. A monitoring and reporting function in conjunction with company level incentives will fill this need; a logic diagram of this monitoring program is given.
Date: July 1, 1996
Creator: Tripp, J.L.
Partner: UNT Libraries Government Documents Department