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Sample records for screening high-level-waste tanks

  1. An approximate-reasoning-based method for screening high-level waste tanks for flammable gas

    International Nuclear Information System (INIS)

    Eisenhawer, S.W.; Bott, T.F.; Smith, R.E.

    1998-01-01

    The in situ retention of flammable gas produced by radiolysis and thermal decomposition in high-level waste can pose a safety problem if the gases are released episodically into the dome space of a storage tank. Screening efforts at Hanford have been directed at identifying tanks in which this situation could exist. Problems encountered in screening motivated an effort to develop an improved screening methodology. Approximate reasoning (AR) is a formalism designed to emulate the kinds of complex judgments made by subject matter experts. It uses inductive logic structures to build a sequence of forward-chaining inferences about a subject. AR models incorporate natural language expressions known as linguistic variables to represent evidence. The use of fuzzy sets to represent these variables mathematically makes it practical to evaluate quantitative and qualitative information consistently. The authors performed a pilot study to investigate the utility of AR for flammable gas screening. They found that the effort to implement such a model was acceptable and that computational requirements were reasonable. The preliminary results showed that important judgments about the validity of observational data and the predictive power of models could be made. These results give new insights into the problems observed in previous screening efforts

  2. An approximate-reasoning-based method for screening high-level waste tanks for flammable gas

    Energy Technology Data Exchange (ETDEWEB)

    Eisenhawer, S.W.; Bott, T.F.; Smith, R.E.

    1998-07-01

    The in situ retention of flammable gas produced by radiolysis and thermal decomposition in high-level waste can pose a safety problem if the gases are released episodically into the dome space of a storage tank. Screening efforts at Hanford have been directed at identifying tanks in which this situation could exist. Problems encountered in screening motivated an effort to develop an improved screening methodology. Approximate reasoning (AR) is a formalism designed to emulate the kinds of complex judgments made by subject matter experts. It uses inductive logic structures to build a sequence of forward-chaining inferences about a subject. AR models incorporate natural language expressions known as linguistic variables to represent evidence. The use of fuzzy sets to represent these variables mathematically makes it practical to evaluate quantitative and qualitative information consistently. The authors performed a pilot study to investigate the utility of AR for flammable gas screening. They found that the effort to implement such a model was acceptable and that computational requirements were reasonable. The preliminary results showed that important judgments about the validity of observational data and the predictive power of models could be made. These results give new insights into the problems observed in previous screening efforts.

  3. Life Extension of Aging High Level Waste (HLW) Tanks

    International Nuclear Information System (INIS)

    The Double Shell Tanks (DSTs) play a critical role in the Hanford High-Level Waste Treatment Complex, and therefore activities are underway to protect and better understand these tanks. The DST Life Extension Program is focused on both tank life extension and on evaluation of tank integrity. Tank life extension activities focus on understanding tank failure modes and have produced key chemistry and operations controls to minimize tank corrosion and extend useful tank life. Tank integrity program activities have developed and applied key technologies to evaluate the condition of the tank structure and predict useful tank life. Program results to date indicate that DST useful life can be extended well beyond the original design life and allow the existing tanks to fill a critical function within the Hanford High-Level Waste Treatment Complex. In addition the tank life may now be more reliably predicted, facilitating improved planning for the use and possible future replacement of these tanks

  4. Nondestructive examination of DOE high-level waste storage tanks

    International Nuclear Information System (INIS)

    Bush, S.; Bandyopadhyay, K.; Kassir, M.; Mather, B.; Shewmon, P.; Streicher, M.; Thompson, B.; van Rooyen, D.; Weeks, J.

    1995-01-01

    A number of DOE sites have buried tanks containing high-level waste. Tanks of particular interest am double-shell inside concrete cylinders. A program has been developed for the inservice inspection of the primary tank containing high-level waste (HLW), for testing of transfer lines and for the inspection of the concrete containment where possible. Emphasis is placed on the ultrasonic examination of selected areas of the primary tank, coupled with a leak-detection system capable of detecting small leaks through the wall of the primary tank. The NDE program is modelled after ASME Section XI in many respects, particularly with respects to the sampling protocol. Selected testing of concrete is planned to determine if there has been any significant degradation. The most probable failure mechanisms are corrosion-related so that the examination program gives major emphasis to possible locations for corrosion attack

  5. Mixing Processes in High-Level Waste Tanks - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, P.F.

    1999-05-24

    The mixing processes in large, complex enclosures using one-dimensional differential equations, with transport in free and wall jets is modeled using standard integral techniques. With this goal in mind, we have constructed a simple, computationally efficient numerical tool, the Berkeley Mechanistic Mixing Model, which can be used to predict the transient evolution of fuel and oxygen concentrations in DOE high-level waste tanks following loss of ventilation, and validate the model against a series of experiments.

  6. Mixing Processes in High-Level Waste Tanks - Final Report

    International Nuclear Information System (INIS)

    Peterson, P.F.

    1999-01-01

    The mixing processes in large, complex enclosures using one-dimensional differential equations, with transport in free and wall jets is modeled using standard integral techniques. With this goal in mind, we have constructed a simple, computationally efficient numerical tool, the Berkeley Mechanistic Mixing Model, which can be used to predict the transient evolution of fuel and oxygen concentrations in DOE high-level waste tanks following loss of ventilation, and validate the model against a series of experiments

  7. Alternatives Generation and Analysis for Heat Removal from High Level Waste Tanks

    Energy Technology Data Exchange (ETDEWEB)

    WILLIS, W.L.

    2000-06-15

    This document addresses the preferred combination of design and operational configurations to provide heat removal from high-level waste tanks during Phase 1 waste feed delivery to prevent the waste temperature from exceeding tank safety requirement limits. An interim decision for the preferred method to remove the heat from the high-level waste tanks during waste feed delivery operations is presented herein.

  8. Alternatives Generation and Analysis for Heat Removal from High Level Waste Tanks

    International Nuclear Information System (INIS)

    WILLIS, W.L.

    2000-01-01

    This document addresses the preferred combination of design and operational configurations to provide heat removal from high-level waste tanks during Phase 1 waste feed delivery to prevent the waste temperature from exceeding tank safety requirement limits. An interim decision for the preferred method to remove the heat from the high-level waste tanks during waste feed delivery operations is presented herein

  9. Decision Document for Heat Removal from High-Level Waste Tanks

    International Nuclear Information System (INIS)

    WILLIS, W.L.

    2000-01-01

    This document establishes the combination of design and operational configurations that will be used to provide heat removal from high-level waste tanks during Phase 1 waste feed delivery to prevent the waste temperature from exceeding tank safety requirement limits. The chosen method--to use the primary and annulus ventilation systems to remove heat from the high-level waste tanks--is documented herein

  10. Guidelines for development of structural integrity programs for DOE high-level waste storage tanks

    Energy Technology Data Exchange (ETDEWEB)

    Bandyopadhyay, K.; Bush, S.; Kassir, M.; Mather, B.; Shewmon, P.; Streicher, M.; Thompson, B.; Rooyen, D. van; Weeks, J.

    1997-01-01

    Guidelines are provided for developing programs to promote the structural integrity of high-level waste storage tanks and transfer lines at the facilities of the Department of Energy. Elements of the program plan include a leak-detection system, definition of appropriate loads, collection of data for possible material and geometric changes, assessment of the tank structure, and non-destructive examination. Possible aging degradation mechanisms are explored for both steel and concrete components of the tanks, and evaluated to screen out nonsignificant aging mechanisms and to indicate methods of controlling the significant aging mechanisms. Specific guidelines for assessing structural adequacy will be provided in companion documents. Site-specific structural integrity programs can be developed drawing on the relevant portions of the material in this document.

  11. 3-D MAPPING TECHNOLOGIES FOR HIGH LEVEL WASTE TANKS

    Energy Technology Data Exchange (ETDEWEB)

    Marzolf, A.; Folsom, M.

    2010-08-31

    This research investigated four techniques that could be applicable for mapping of solids remaining in radioactive waste tanks at the Savannah River Site: stereo vision, LIDAR, flash LIDAR, and Structure from Motion (SfM). Stereo vision is the least appropriate technique for the solids mapping application. Although the equipment cost is low and repackaging would be fairly simple, the algorithms to create a 3D image from stereo vision would require significant further development and may not even be applicable since stereo vision works by finding disparity in feature point locations from the images taken by the cameras. When minimal variation in visual texture exists for an area of interest, it becomes difficult for the software to detect correspondences for that object. SfM appears to be appropriate for solids mapping in waste tanks. However, equipment development would be required for positioning and movement of the camera in the tank space to enable capturing a sequence of images of the scene. Since SfM requires the identification of distinctive features and associates those features to their corresponding instantiations in the other image frames, mockup testing would be required to determine the applicability of SfM technology for mapping of waste in tanks. There may be too few features to track between image frame sequences to employ the SfM technology since uniform appearance may exist when viewing the remaining solids in the interior of the waste tanks. Although scanning LIDAR appears to be an adequate solution, the expense of the equipment ($80,000-$120,000) and the need for further development to allow tank deployment may prohibit utilizing this technology. The development would include repackaging of equipment to permit deployment through the 4-inch access ports and to keep the equipment relatively uncontaminated to allow use in additional tanks. 3D flash LIDAR has a number of advantages over stereo vision, scanning LIDAR, and SfM, including full frame

  12. Potential for erosion corrosion of SRS high level waste tanks

    International Nuclear Information System (INIS)

    Zapp, P.E.

    1994-01-01

    SRS high-level radioactive waste tanks will not experience erosion corrosion to any significant degree during slurry pump operations. Erosion corrosion in carbon steel structures at reported pump discharge velocities is dominated by electrochemical (corrosion) processes. Interruption of those processes, as by the addition of corrosion inhibitors, sharply reduces the rate of metal loss from erosion corrosion. The well-inhibited SRS waste tanks have a near-zero general corrosion rate, and therefore will be essentially immune to erosion corrosion. The experimental data on carbon steel erosion corrosion most relevant to SRS operations was obtained at the Hanford Site on simulated Purex waste. A metal loss rate of 2.4 mils per year was measured at a temperature of 102 C and a slurry velocity comparable to calculated SRS slurry velocities on ground specimens of the same carbon steel used in SRS waste tanks. Based on these data and the much lower expected temperatures, the metal loss rate of SRS tanks under waste removal and processing conditions should be insignificant, i.e. less than 1 mil per year

  13. High-level waste tank farm set point document

    International Nuclear Information System (INIS)

    Anthony, J.A. III.

    1995-01-01

    Setpoints for nuclear safety-related instrumentation are required for actions determined by the design authorization basis. Minimum requirements need to be established for assuring that setpoints are established and held within specified limits. This document establishes the controlling methodology for changing setpoints of all classifications. The instrumentation under consideration involve the transfer, storage, and volume reduction of radioactive liquid waste in the F- and H-Area High-Level Radioactive Waste Tank Farms. The setpoint document will encompass the PROCESS AREA listed in the Safety Analysis Report (SAR) (DPSTSA-200-10 Sup 18) which includes the diversion box HDB-8 facility. In addition to the PROCESS AREAS listed in the SAR, Building 299-H and the Effluent Transfer Facility (ETF) are also included in the scope

  14. High-level waste tank farm set point document

    Energy Technology Data Exchange (ETDEWEB)

    Anthony, J.A. III

    1995-01-15

    Setpoints for nuclear safety-related instrumentation are required for actions determined by the design authorization basis. Minimum requirements need to be established for assuring that setpoints are established and held within specified limits. This document establishes the controlling methodology for changing setpoints of all classifications. The instrumentation under consideration involve the transfer, storage, and volume reduction of radioactive liquid waste in the F- and H-Area High-Level Radioactive Waste Tank Farms. The setpoint document will encompass the PROCESS AREA listed in the Safety Analysis Report (SAR) (DPSTSA-200-10 Sup 18) which includes the diversion box HDB-8 facility. In addition to the PROCESS AREAS listed in the SAR, Building 299-H and the Effluent Transfer Facility (ETF) are also included in the scope.

  15. Criticality Safety Evaluation of Hanford Site High-Level Waste Storage Tanks

    International Nuclear Information System (INIS)

    ROGERS, C.A.

    2000-01-01

    This criticality safety evaluation covers operations for waste in underground storage tanks at the high-level waste tank farms on the Hanford site. This evaluation provides the bases for criticality safety limits and controls to govern receipt, transfer, and long-term storage of tank waste. Justification is provided that a nuclear criticality accident cannot occur for tank farms operations, based on current fissile material and operating conditions

  16. Criticality Safety Evaluation of Hanford Site High Level Waste Storage Tanks

    Energy Technology Data Exchange (ETDEWEB)

    ROGERS, C.A.

    2000-02-17

    This criticality safety evaluation covers operations for waste in underground storage tanks at the high-level waste tank farms on the Hanford site. This evaluation provides the bases for criticality safety limits and controls to govern receipt, transfer, and long-term storage of tank waste. Justification is provided that a nuclear criticality accident cannot occur for tank farms operations, based on current fissile material and operating conditions.

  17. Tank waste remediation system phase I high-level waste feed processability assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Lambert, S.L.; Stegen, G.E., Westinghouse Hanford

    1996-08-01

    This report evaluates the effects of feed composition on the Phase I high-level waste immobilization process and interim storage facility requirements for the high-level waste glass.Several different Phase I staging (retrieval, blending, and pretreatment) scenarios were used to generate example feed compositions for glass formulations, testing, and glass sensitivity analysis. Glass models and data form laboratory glass studies were used to estimate achievable waste loading and corresponding glass volumes for various Phase I feeds. Key issues related to feed process ability, feed composition, uncertainty, and immobilization process technology are identified for future consideration in other tank waste disposal program activities.

  18. DELPHI expert panel evaluation of Hanford high level waste tank failure modes and release quantities

    Energy Technology Data Exchange (ETDEWEB)

    Dunford, G.L.; Han, F.C.

    1996-09-30

    The Failure Modes and Release Quantities of the Hanford High Level Waste Tanks due to postulated accident loads were established by a DELPHI Expert Panel consisting of both on-site and off-site experts in the field of Structure and Release. The Report presents the evaluation process, accident loads, tank structural failure conclusion reached by the panel during the two-day meeting.

  19. HIGH LEVEL WASTE TANK CLOSURE PROJECT AT THE IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LABORATORY

    International Nuclear Information System (INIS)

    Quigley, K.D.; Wessman, D.

    2003-01-01

    The Department of Energy, Idaho Operations Office (DOE-ID) is in the process of closing two underground high-level waste (HLW) storage tanks at the Idaho National Engineering and Environmental Laboratory (INEEL) to meet Resource Conservation and Recovery Act (RCRA) regulations and Department of Energy orders. Closure of these two tanks is scheduled for 2004 as the first phase in closure of the eleven 1.14 million liter (300,000 gallon) tanks currently in service at the Idaho Nuclear Technology and Engineering Center (INTEC). The INTEC Tank Farm Facility (TFF) Closure sequence consists of multiple steps to be accomplished through the existing tank riser access points. Currently, the tank risers contain steam and process waste lines associated with the steam jets, corrosion coupons, and liquid level indicators. As necessary, this equipment will be removed from the risers to allow adequate space for closure equipment and activities. The basic tank closure sequence is as follows: Empty the tank to the residual heel using the existing jets; Video and sample the heel; Replace steam jets with new jet at a lower position in the tank, and remove additional material; Flush tank, piping and secondary containment with demineralized water; Video and sample the heel; Evaluate decontamination effectiveness; Displace the residual heel with multiple placements of grout; and Grout piping, vaults and remaining tank volume. Design, development, and deployment of a remotely operated tank cleaning system were completed in June 2002. The system incorporates many commercially available components, which have been adapted for application in cleaning high-level waste tanks. The system is cost-effective since it also utilizes existing waste transfer technology (steam jets), to remove tank heel solids from the tank bottoms during the cleaning operations. Remotely operated directional spray nozzles, automatic rotating wash balls, video monitoring equipment, decontamination spray-rings, and

  20. M.A. Streicher findings regarding high-level waste tank corrosion issues

    International Nuclear Information System (INIS)

    Husa, E.I.

    1994-01-01

    Dr. Michael A. Streicher is a nationally recognized metallurgist and corrosion scientist. He has served on the Department of Energy, Headquarters Tank Structural Integrity panel as the primary corrosion technical expert since the panel's inception in October 1991. Attachments 3 through 13 are Dr. Streicher's correspondence and presentations to the panel between November 1991 and May 1994. This compilation addresses Dr. Streicher's findings on High-Level Waste tank corrosion issues such as: corrosion mechanisms in carbon steels; hydrogen generation from waste tank corrosion; stress corrosion cracking in carbon steel tanks; water line attack in Hanford's single-shell tanks; stress corrosion cracking of austenitic stainless steels; and materials selection for new Hanford waste tanks. These papers discuss both generic and specific corrosion issues associated with waste tanks and transfer systems at Hanford, Savannah River, Idaho National Engineering Laboratory, and West Valley Demonstration Project

  1. High-level waste tank modifications, installation of mobilization equipment/check out

    International Nuclear Information System (INIS)

    Schiffhauer, M.A.; Thompson, S.C.

    1992-01-01

    PUREX high-level waste (HLW) is contained at the West Valley Demonstration Project (WVDP) in an underground carbon-steel storage tank. The HLW consists of a precipitated sludge and an alkaline supernate. This report describes the system that the WVDP has developed and implemented to resuspend and wash the HLW sludge from the tank. The report discusses Sludge Mobilization and Wash System (SMWS) equipment design, installation, and testing. The storage tank required modifications to accommodate the SMWS. These modifications are discussed as well

  2. The remote installation of a new riser on an existing high level waste (HLW) tank

    International Nuclear Information System (INIS)

    Scott, D.W.; Moore, T.L.; Schiffhauer, M.A.

    1985-01-01

    On October 1, 1980, the U.S. Congress passed the WVDP Act, directing the U.S. Department of Energy (DOE) to undertake a high level radioactive waste management demonstration project at the West Valley Site. Under the Act, the DOE is responsible for solidifying the high level waste in a form suitable for transportation to a federal repository for ultimate disposal. West Valley Nuclear Services Company, Inc. (WVNS), a wholly-owned subsidiary of Westinghouse Electric Corporation, is implementing the Project for the DOE. The largest volume of high level waste (HLW) stored underground at West Valley is in Tank 8D-2, containing approximately 2,250 m/sup 3/ of supernatant and 170 m/sup 3/ of sludge. An identical spare tank (8D-1) is approximately one-third full of contaminated condensate. The supernatant phase of the waste will be decontaminated to a level which will permit it to be solidified in concrete and disposed of as low level waste. The radioactive contaminants removed from the supernatant will be combined with the balance of the high level waste, delivered to a slurry fed ceramic melter (SFCM), and solidified into borosilicate glass. The waste must first be removed from these tanks before it can be vitrified. This paper addresses the task of gaining access to the tank interiors for the removal process. The complex internal geometry of the tank requires a special array of sluicing pumps to homogenize the supernatant and sludge for the removal process. The one existing riser large enough to install a slurry pump is insufficient to provide adequate removal of the waste

  3. Overview of Hanford Site High-Level Waste Tank Gas and Vapor Dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Huckaby, James L.; Mahoney, Lenna A.; Droppo, James G.; Meacham, Joseph E.

    2004-08-31

    Hanford Site processes associated with the chemical separation of plutonium from uranium and other fission products produced a variety of volatile, semivolatile, and nonvolatile organic and inorganic waste chemicals that were sent to high-level waste tanks. These chemicals have undergone and continue to undergo radiolytic and thermal reactions in the tanks to produce a wide variety of degradation reaction products. The origins of the organic wastes, the chemical reactions they undergo, and their reaction products have recently been examined by Stock (2004). Stock gives particular attention to explaining the presence of various types of volatile and semivolatile organic species identified in headspace air samples. This report complements the Stock report by examining the storage of volatile and semivolatile species in the waste, their transport through any overburden of waste to the tank headspaces, the physical phenomena affecting their concentrations in the headspaces, and their eventual release into the atmosphere above the tanks.

  4. Resolution of the ferrocyanide safety issue for the Hanford site high-level waste tanks

    International Nuclear Information System (INIS)

    Cash, R.J.

    1996-01-01

    This paper describes the approach used to resolve the ferrocyanide safety issue, a process that began in 1990 after heightened concern was expressed by various government agencies about the safety of Hanford site high-level waste tanks. At the time, little was known about ferrocyanide-nitrate/nitrite reactions and the potential for offsite releases of radioactivity from the Hanford Site. Recent studies have shown that the combined effects of temperature, radiation, and pH during more than 38 years of storage have destroyed most of the ferrocyanide originally added to tanks. This has been proven in the laboratory using flowsheet-derived waste simulants and confirmed by waste samples obtained from the ferrocyanide tanks. The resulting tank waste sludges are too dilute to support a sustained exothermic reaction, even if dried out and heated to temperatures of at least 250 C. The US Department of Energy (DOE) has been requested to close the ferrocyanide safety issue

  5. Annual Report, Fall 2016: Alternative Chemical Cleaning of Radioactive High Level Waste Tanks - Corrosion Test Results

    Energy Technology Data Exchange (ETDEWEB)

    Wyrwas, R. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-01

    The testing presented in this report is in support of the investigation of the Alternative Chemical Cleaning program to aid in developing strategies and technologies to chemically clean radioactive High Level Waste tanks prior to tank closure. The data and conclusions presented here were the examination of the corrosion rates of A285 carbon steel and 304L stainless steel exposed to two proposed chemical cleaning solutions: acidic permanganate (0.18 M nitric acid and 0.05M sodium permanganate) and caustic permanganate. (10 M sodium hydroxide and 0.05M sodium permanganate). These solutions have been proposed as a chemical cleaning solution for the retrieval of actinides in the sludge in the waste tanks, and were tested with both HM and PUREX sludge simulants at a 20:1 ratio.

  6. Modeling Carbonation of High-Level Waste Tank Integrity and Closure

    Directory of Open Access Journals (Sweden)

    Meeussen J.C.L.

    2013-07-01

    Full Text Available The Cementitious Barriers Partnership (CBP is focused on reducinguncertainties in current methodologies for assessing cementitious barrier performanceand increasing the consistency and transparency in the assessment process. Oneimportant set of US Department of Energy challenges is assessing the integrity andclosure of the high-level waste (HLW tanks that currently store millions of gallons ofhighly radioactive wastes. Many of these tanks are decades past their design lives, haveleaked or been overfilled, and must be emptied and closed to satisfy regulatoryagreements. Carbonation-induced corrosion has been identified as a primary degradationand possible failure mechanism for the HLW tanks prior to closure. After closure theimpact of carbonation (and concurrent oxidation may be to increase the release andshort-range transport of contaminants of concern. HLW tanks may be significantlyempty for many years (and possibly decades prior to closure; the performance of theclosed tank over centuries, if not millennia, must be assessed to evaluate the potentialrelease of residual radionuclides to the environment. CBP is developing models to evaluate a representative HLW tank closure scenarioincluding the potential impacts of carbonation on waste tanks prior to and post closure.CBP modeling tools, including LeachXS™/ORCHESTRA, are being used to simulatewaste tank carbonation, major constituent leaching, and contaminant releases to evaluatethe source term and near-field conditions. Simulations presented here include sensitivityanalysis for uncracked concrete to varying input parameters including composition,effective diffusivities, and thermodynamic parameters.

  7. Alternative Chemical Cleaning Methods for High Level Waste Tanks: Simulant Studies

    Energy Technology Data Exchange (ETDEWEB)

    Rudisill, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); King, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hay, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jones, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-11-19

    Solubility testing with simulated High Level Waste tank heel solids has been conducted in order to evaluate two alternative chemical cleaning technologies for the dissolution of sludge residuals remaining in the tanks after the exhaustion of mechanical cleaning and sludge washing efforts. Tests were conducted with non-radioactive pure phase metal reagents, binary mixtures of reagents, and a Savannah River Site PUREX heel simulant to determine the effectiveness of an optimized, dilute oxalic/nitric acid cleaning reagent and pure, dilute nitric acid toward dissolving the bulk non-radioactive waste components. A focus of this testing was on minimization of oxalic acid additions during tank cleaning. For comparison purposes, separate samples were also contacted with pure, concentrated oxalic acid which is the current baseline chemical cleaning reagent. In a separate study, solubility tests were conducted with radioactive tank heel simulants using acidic and caustic permanganate-based methods focused on the “targeted” dissolution of actinide species known to be drivers for Savannah River Site tank closure Performance Assessments. Permanganate-based cleaning methods were evaluated prior to and after oxalic acid contact.

  8. Aging mechanisms for concrete components of High-Level Waste storage tanks

    International Nuclear Information System (INIS)

    Kassir, M.; Bandyopadhyay, K.; Bush, S.; Mather, B.; Shewmon, P.; Streicher, M.; Thompson, B.; van Rooyen, D.; Weeks, J.

    1995-01-01

    The age-related degradation mechanisms which affect the concrete and the reinforcing steel in the high-level waste (HLW) storage tanks art evaluated with respect to their potential significance to the continued performance of the concrete, and am classified into non-significant and potentially significant. The identified potentially significant degradation mechanisms include the effects of elevated temperature, freezing and thawing, leaching of calcium hydroxide, aggressive chemical attack, and corrosion of the reinforcing steel. To the extent that available knowledge permits, these mechanisms are generically evaluated and quantified so that site-specific plans may be developed to verify whether significant degradation has occurred in the concrete, and, if so, to formulate mitigating measures to avoid further deterioration and possibly repair the degradation or pursue other management options

  9. ESTIMATING HIGH LEVEL WASTE MIXING PERFORMANCE IN HANFORD DOUBLE SHELL TANKS

    International Nuclear Information System (INIS)

    Thien, M.G.; Greer, D.A.; Townson, P.

    2011-01-01

    The ability to effectively mix, sample, certify, and deliver consistent batches of high level waste (HLW) feed from the Hanford double shell tanks (DSTs) to the Waste Treatment and Immobilization Plant (WTP) presents a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. The Department of Energy's (DOE's) Tank Operations Contractor (TOC), Washington River Protection Solutions (WRPS) is currently demonstrating mixing, sampling, and batch transfer performance in two different sizes of small-scale DSTs. The results of these demonstrations will be used to estimate full-scale DST mixing performance and provide the key input to a programmatic decision on the need to build a dedicated feed certification facility. This paper discusses the results from initial mixing demonstration activities and presents data evaluation techniques that allow insight into the performance relationships of the two small tanks. The next steps, sampling and batch transfers, of the small scale demonstration activities are introduced. A discussion of the integration of results from the mixing, sampling, and batch transfer tests to allow estimating full-scale DST performance is presented.

  10. Mixing processes in high-level waste tanks. Progress report, September 15, 1996 - September 14, 1997

    International Nuclear Information System (INIS)

    Peterson, P.F.

    1997-01-01

    'U.C. Berkeley has made excellent progress in the last year in building and running experiments and performing analysis to study mixing processes that can affect the distribution of fuel and oxygen in the air space of DOE high-level waste tanks, and the potential to create flammable concentrations at isolated locations, achieving all of the milestones outlined in the proposal. The DOE support has allowed the acquisition of key experimental equipment, and has funded the full-time efforts of one doctoral student and one postdoctoral researcher working on the project. In addition, one masters student and one other doctoral student, funded by external sources, have also contributed to the research effort. Flammable gases can be generated in DOE high-level waste tanks, including radiolytic hydrogen, and during cesium precipitation from salt solutions, benzene. Under normal operating conditions the potential for deflagration or detonation from these gases is precluded by purging and ventilation systems, which remove the flammable gases and maintain a well-mixed condition in the tanks. Upon failure of the ventilation system, due to seismic or other events, however, it has proven more difficult to make strong arguments for well-mixed conditions, due to the potential for density-induced stratification which can potentially sequester fuel or oxidizer at concentrations significantly higher than average. This has complicated the task of defining the safety basis for tank operation. The author is currently developing numerical tools for modeling the transient evolution of fuel and oxygen concentrations in waste tanks following loss of ventilation. When used with reasonable grid resolutions, standard multi-dimensional fluid dynamics codes suffer from excessive numerical diffusion effects, which strongly over predict mixing and provide nonconservative estimates, particularly after stratification occurs. The National Institute of Standards and Technology (NIST) has developed

  11. Derivation of residual radionuclide inventory guidelines for implace closure of high-level waste tanks

    International Nuclear Information System (INIS)

    Yuan, L.; Yuan, Y.

    1999-01-01

    Residual radionuclide inventory guidelines were derived for the high-level waste tanks at a vitrification facility. The decommissioning scenario assumed for this derivation was that the tanks were to be stabilized at the present locations and the site is released for unrestricted use following a 100-year institutional control period. It was assumed that loss of institutional control would occur at 100-years following tank closure. The derivation of the residual radionuclide inventory guidelines was based on the requirement that the effective dose equivalent (EDE) to a hypothetical individual who lives in the vicinity of the site should not exceed a dose of 0.15 mSv/yr off-site and 5 mSv/yr on-site following closure of the tanks. The RESRAD computer code, modified for exposure scenarios specific for the site, was used for this evaluation. The results of the derivation indicate that the allowable off-site dose limit will not be exceeded. The estimated potential doses to individuals using water offsite from a creek are negligibly small fractions of the 0.15 mSv/yr allowable dose limit. With an assumed 3% heel remaining in the tanks, the estimated peak dose rate for the future offsite water user is about 0.00025 mSv/yr. The residual radionuclide inventory guidelines derived based on potential doses to the on-site resident farmer indicate that, with the exception of Tc-99 and C-14, a 3% heel remaining in the tanks would not result in doses exceeding the 5 mSv/yr allowable dose limit. For this on-site exposure scenario, the peak dose rates occur at about 2000 years after tank closure. The peak dose rate is calculated to be 25 mSv/yr, with greater than 99% produced by four radionuclides: C-14, Tc-99, Np-237, and Am-241. Ingestion of contaminated vegetation contributes most (90%) of the peak dose. Since the inventories used for the derivation are mostly estimated from fuel depletion calculations. There is a need to determine further the actual inventories of these

  12. DOUBLE SHELL TANK (DST) INTEGRITY PROJECT HIGH LEVEL WASTE CHEMISTRY OPTIMIZATION

    Energy Technology Data Exchange (ETDEWEB)

    WASHENFELDER DJ

    2008-01-22

    The U.S. Department of Energy's Office (DOE) of River Protection (ORP) has a continuing program for chemical optimization to better characterize corrosion behavior of High-Level Waste (HLW). The DOE controls the chemistry in its HLW to minimize the propensity of localized corrosion, such as pitting, and stress corrosion cracking (SCC) in nitrate-containing solutions. By improving the control of localized corrosion and SCC, the ORP can increase the life of the Double-Shell Tank (DST) carbon steel structural components and reduce overall mission costs. The carbon steel tanks at the Hanford Site are critical to the mission of safely managing stored HLW until it can be treated for disposal. The DOE has historically used additions of sodium hydroxide to retard corrosion processes in HLW tanks. This also increases the amount of waste to be treated. The reactions with carbon dioxide from the air and solid chemical species in the tank continually deplete the hydroxide ion concentration, which then requires continued additions. The DOE can reduce overall costs for caustic addition and treatment of waste, and more effectively utilize waste storage capacity by minimizing these chemical additions. Hydroxide addition is a means to control localized and stress corrosion cracking in carbon steel by providing a passive environment. The exact mechanism that causes nitrate to drive the corrosion process is not yet clear. The SCC is less of a concern in the newer stress relieved double shell tanks due to reduced residual stress. The optimization of waste chemistry will further reduce the propensity for SCC. The corrosion testing performed to optimize waste chemistry included cyclic potentiodynamic volarization studies. slow strain rate tests. and stress intensity factor/crack growth rate determinations. Laboratory experimental evidence suggests that nitrite is a highly effective:inhibitor for pitting and SCC in alkaline nitrate environments. Revision of the corrosion control

  13. A Dual Regime Reactive Transport Model for Simulation of High Level Waste Tank Closure Scenarios - 13375

    International Nuclear Information System (INIS)

    Sarkar, Sohini; Kosson, David S.; Brown, Kevin; Garrabrants, Andrew C.; Meeussen, Hans; Van der Sloot, Hans

    2013-01-01

    A numerical simulation framework is presented in this paper for estimating evolution of pH and release of major species from grout within high-level waste tanks after closure. This model was developed as part of the Cementitious Barriers Partnership. The reactive transport model consists of two parts - (1) transport of species, and (2) chemical reactions. The closure grout can be assumed to have varying extents of cracking and composition for performance assessment purposes. The partially or completely degraded grouted tank is idealized as a dual regime system comprising of a mobile region having solid materials with cracks and macro-pores, and an immobile/stagnant region having solid matrix with micropores. The transport profiles of the species are calculated by incorporating advection of species through the mobile region, diffusion of species through the immobile/stagnant region, and exchange of species between the mobile and immobile regions. A geochemical speciation code in conjunction with the pH dependent test data for a grout material is used to obtain a mineral set that best describes the trends in the test data of the major species. The dual regime reactive transport model predictions are compared with the release data from an up-flow column percolation test. The coupled model is then used to assess effects of crack state of the structure, rate and composition of the infiltrating water on the pH evolution at the grout-waste interface. The coupled reactive transport model developed in this work can be used as part of the performance assessment process for evaluating potential risks from leaching of a cracked tank containing elements of human health and environmental concern. (authors)

  14. High level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 6

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 6) outlines the standards and requirements for the sections on: Environmental Restoration and Waste Management, Research and Development and Experimental Activities, and Nuclear Safety.

  15. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID)

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 3) presents the standards and requirements for the following sections: Safeguards and Security, Engineering Design, and Maintenance.

  16. High level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 6

    International Nuclear Information System (INIS)

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 6) outlines the standards and requirements for the sections on: Environmental Restoration and Waste Management, Research and Development and Experimental Activities, and Nuclear Safety

  17. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID)

    International Nuclear Information System (INIS)

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 3) presents the standards and requirements for the following sections: Safeguards and Security, Engineering Design, and Maintenance

  18. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Document (S/RID) is contained in multiple volumes. This document (Volume 2) presents the standards and requirements for the following sections: Quality Assurance, Training and Qualification, Emergency Planning and Preparedness, and Construction.

  19. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 4

    International Nuclear Information System (INIS)

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 4) presents the standards and requirements for the following sections: Radiation Protection and Operations

  20. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 5

    International Nuclear Information System (INIS)

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 5) outlines the standards and requirements for the Fire Protection and Packaging and Transportation sections

  1. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 2

    International Nuclear Information System (INIS)

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Document (S/RID) is contained in multiple volumes. This document (Volume 2) presents the standards and requirements for the following sections: Quality Assurance, Training and Qualification, Emergency Planning and Preparedness, and Construction

  2. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 4

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 4) presents the standards and requirements for the following sections: Radiation Protection and Operations.

  3. Demonstration of Small Tank Tetraphenylborate Precipitation Process Using Savannah River Site High Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    Peters, T.B.

    2001-09-10

    This report details the experimental effort to demonstrate the continuous precipitation of cesium from Savannah River Site High Level Waste using sodium tetraphenylborate. In addition, the experiments examined the removal of strontium and various actinides through addition of monosodium titanate.

  4. Alternative Chemical Cleaning Methods for High Level Waste Tanks: Actual Waste Testing with SRS Tank 5F Sludge

    Energy Technology Data Exchange (ETDEWEB)

    King, William D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hay, Michael S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-30

    Solubility testing with actual High Level Waste tank sludge has been conducted in order to evaluate several alternative chemical cleaning technologies for the dissolution of sludge residuals remaining in the tanks after the exhaustion of mechanical cleaning and sludge sluicing efforts. Tests were conducted with archived Savannah River Site (SRS) radioactive sludge solids that had been retrieved from Tank 5F in order to determine the effectiveness of an optimized, dilute oxalic/nitric acid cleaning reagent toward dissolving the bulk non-radioactive waste components. Solubility tests were performed by direct sludge contact with the oxalic/nitric acid reagent and with sludge that had been pretreated and acidified with dilute nitric acid. For comparison purposes, separate samples were also contacted with pure, concentrated oxalic acid following current baseline tank chemical cleaning methods. One goal of testing with the optimized reagent was to compare the total amounts of oxalic acid and water required for sludge dissolution using the baseline and optimized cleaning methods. A second objective was to compare the two methods with regard to the dissolution of actinide species known to be drivers for SRS tank closure Performance Assessments (PA). Additionally, solubility tests were conducted with Tank 5 sludge using acidic and caustic permanganate-based methods focused on the “targeted” dissolution of actinide species.

  5. Implementation of seismic design and evaluation guidelines for the Department of Energy high-level waste storage tanks and appurtenances

    International Nuclear Information System (INIS)

    Conrads, T.J.

    1993-06-01

    In the fall of 1992, a draft of the Seismic Design and Evaluation Guidelines for the Department of Energy (DOE) High-level Waste Storage Tanks and Appurtenances was issued. The guidelines were prepared by the Tanks Seismic Experts Panel (TSEP) and this task was sponsored by DOE, Environmental Management. The TSEP is comprised of a number of consultants known for their knowledge of seismic ground motion and expertise in the analysis of structures, systems and components subjected to seismic loads. The development of these guidelines was managed by staff from Brookhaven National Laboratory, Engineering Research and Applications Division, Department of Nuclear Energy. This paper describes the process used to incorporate the Seismic Design and Evaluation Guidelines for the DOE High-Level Waste Storage Tanks and Appurtenances into the design criteria for the Multi-Function Waste Tank Project at the Hanford Site. This project will design and construct six new high-level waste tanks in the 200 Areas at the Hanford Site. This paper also discusses the vehicles used to ensure compliance to these guidelines throughout Title 1 and Title 2 design phases of the project as well as the strategy used to ensure consistent and cost-effective application of the guidelines by the structural analysts. The paper includes lessons learned and provides recommendations for other tank design projects which might employ the TSEP guidelines

  6. THE RETRIEVAL KNOWLEDGE CENTER EVALUATION OF LOW TANK LEVEL MIXING TECHNOLOGIES FOR DOE HIGH LEVEL WASTE TANK RETRIEVAL 10516

    Energy Technology Data Exchange (ETDEWEB)

    Fellinger, A.

    2009-12-08

    The Department of Energy (DOE) Complex has over two-hundred underground storage tanks containing over 80-million gallons of legacy waste from the production of nuclear weapons. The majority of the waste is located at four major sites across the nation and is planned for treatment over a period of almost forty years. The DOE Office of Technology Innovation & Development within the Office of Environmental Management (DOE-EM) sponsors technology research and development programs to support processing advancements and technology maturation designed to improve the costs and schedule for disposal of the waste and closure of the tanks. Within the waste processing focus area are numerous technical initiatives which included the development of a suite of waste removal technologies to address the need for proven equipment and techniques to remove high level radioactive wastes from the waste tanks that are now over fifty years old. In an effort to enhance the efficiency of waste retrieval operations, the DOE-EM Office of Technology Innovation & Development funded an effort to improve communications and information sharing between the DOE's major waste tank locations as it relates to retrieval. The task, dubbed the Retrieval Knowledge Center (RKC) was co-lead by the Savannah River National Laboratory (SRNL) and the Pacific Northwest National Laboratory (PNNL) with core team members representing the Oak Ridge and Idaho sites, as well as, site contractors responsible for waste tank operations. One of the greatest challenges to the processing and closure of many of the tanks is complete removal of all tank contents. Sizeable challenges exist for retrieving waste from High Level Waste (HLW) tanks; with complications that are not normally found with tank retrieval in commercial applications. Technologies currently in use for waste retrieval are generally adequate for bulk removal; however, removal of tank heels, the materials settled in the bottom of the tank, using the same

  7. Seismic design and evaluation guidelines for the Department of Energy high-level waste storage tanks and appurtenances

    International Nuclear Information System (INIS)

    Bandyopadhyay, K.; Cornell, A.; Costantino, C.; Kennedy, R.; Miller, C.; Veletsos, A.

    1993-01-01

    This document provides guidelines for the design and evaluation of underground high-level waste storage tanks due to seismic loads. Attempts were made to reflect the knowledge acquired in the last two decades in the areas of defining the ground motion and calculating hydrodynamic loads and dynamic soil pressures for underground tank structures. The application of the analysis approach is illustrated with an example. The guidelines are developed for specific design of underground storage tanks, namely double-shell structures. However, the methodology discussed is applicable for other types of tank structures as well. The application of these and of suitably adjusted versions of these concepts to other structural types will be addressed in a future version of this document

  8. Tank waste remediation system high-level waste vitrification system development and testing requirements

    International Nuclear Information System (INIS)

    Calmus, R.B.

    1995-01-01

    This document provides the fiscal year (FY) 1995 recommended high-level waste melter system development and testing (D and T) requirements. The first phase of melter system testing (FY 1995) will focus on the feasibility of high-temperature operation of recommended high-level waste melter systems. These test requirements will be used to establish the basis for defining detailed testing work scope, cost, and schedules. This document includes a brief summary of the recommended technologies and technical issues associated with each technology. In addition, this document presents the key D and T activities and engineering evaluations to be performed for a particular technology or general melter system support feature. The strategy for testing in Phase 1 (FY 1995) is to pursue testing of the recommended high-temperature technologies, namely the high-temperature, ceramic-lined, joule-heated melter, referred to as the HTCM, and the high-frequency, cold-wall, induction-heated melter, referred to as the cold-crucible melter (CCM). This document provides a detailed description of the FY 1995 D and T needs and requirements relative to each of the high-temperature technologies

  9. HIGH LEVEL WASTE MECHANCIAL SLUDGE REMOVAL AT THE SAVANNAH RIVER SITE F TANK FARM CLOSURE PROJECT

    International Nuclear Information System (INIS)

    Jolly, R; Bruce Martin, B

    2008-01-01

    The Savannah River Site F-Tank Farm Closure project has successfully performed Mechanical Sludge Removal (MSR) using the Waste on Wheels (WOW) system for the first time within one of its storage tanks. The WOW system is designed to be relatively mobile with the ability for many components to be redeployed to multiple waste tanks. It is primarily comprised of Submersible Mixer Pumps (SMPs), Submersible Transfer Pumps (STPs), and a mobile control room with a control panel and variable speed drives. In addition, the project is currently preparing another waste tank for MSR utilizing lessons learned from this previous operational activity. These tanks, designated as Tank 6 and Tank 5 respectively, are Type I waste tanks located in F-Tank Farm (FTF) with a capacity of 2,840 cubic meters (750,000 gallons) each. The construction of these tanks was completed in 1953, and they were placed into waste storage service in 1959. The tank's primary shell is 23 meters (75 feet) in diameter, and 7.5 meters (24.5 feet) in height. Type I tanks have 34 vertically oriented cooling coils and two horizontal cooling coil circuits along the tank floor. Both Tank 5 and Tank 6 received and stored F-PUREX waste during their operating service time before sludge removal was performed. DOE intends to remove from service and operationally close (fill with grout) Tank 5 and Tank 6 and other HLW tanks that do not meet current containment standards. Mechanical Sludge Removal, the first step in the tank closure process, will be followed by chemical cleaning. After obtaining regulatory approval, the tanks will be isolated and filled with grout for long-term stabilization. Mechanical Sludge Removal operations within Tank 6 removed approximately 75% of the original 95,000 liters (25,000 gallons). This sludge material was transferred in batches to an interim storage tank to prepare for vitrification. This operation consisted of eleven (11) Submersible Mixer Pump(s) mixing campaigns and multiple intraarea

  10. HIGH LEVEL WASTE MECHANCIAL SLUDGE REMOVAL AT THE SAVANNAH RIVER SITE F TANK FARM CLOSURE PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    Jolly, R; Bruce Martin, B

    2008-01-15

    The Savannah River Site F-Tank Farm Closure project has successfully performed Mechanical Sludge Removal (MSR) using the Waste on Wheels (WOW) system for the first time within one of its storage tanks. The WOW system is designed to be relatively mobile with the ability for many components to be redeployed to multiple waste tanks. It is primarily comprised of Submersible Mixer Pumps (SMPs), Submersible Transfer Pumps (STPs), and a mobile control room with a control panel and variable speed drives. In addition, the project is currently preparing another waste tank for MSR utilizing lessons learned from this previous operational activity. These tanks, designated as Tank 6 and Tank 5 respectively, are Type I waste tanks located in F-Tank Farm (FTF) with a capacity of 2,840 cubic meters (750,000 gallons) each. The construction of these tanks was completed in 1953, and they were placed into waste storage service in 1959. The tank's primary shell is 23 meters (75 feet) in diameter, and 7.5 meters (24.5 feet) in height. Type I tanks have 34 vertically oriented cooling coils and two horizontal cooling coil circuits along the tank floor. Both Tank 5 and Tank 6 received and stored F-PUREX waste during their operating service time before sludge removal was performed. DOE intends to remove from service and operationally close (fill with grout) Tank 5 and Tank 6 and other HLW tanks that do not meet current containment standards. Mechanical Sludge Removal, the first step in the tank closure process, will be followed by chemical cleaning. After obtaining regulatory approval, the tanks will be isolated and filled with grout for long-term stabilization. Mechanical Sludge Removal operations within Tank 6 removed approximately 75% of the original 95,000 liters (25,000 gallons). This sludge material was transferred in batches to an interim storage tank to prepare for vitrification. This operation consisted of eleven (11) Submersible Mixer Pump(s) mixing campaigns and multiple

  11. Probabilistic safety assessment for Hanford high-level waste tank 241-SY-101

    International Nuclear Information System (INIS)

    MacFarlane, D.R.; Bott, T.F.; Brown, L.F.; Stack, D.W.; Kindinger, J.; Deremer, R.K.; Medhekar, S.R.; Mikschl, T.J.

    1994-05-01

    Los Alamos National Laboratory (Los Alamos) is performing a comprehensive probabilistic safety assessment (PSA), which will include consideration of external events for the 18 tank farms at the Hanford Site. This effort is sponsored by the Department of Energy (DOE/EM, EM-36). Even though the methodology described herein will be applied to the entire tank farm, this report focuses only on the risk from the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases (open-quotes burpsclose quotes) a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed first because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is being conducted in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. At the Hanford Site there are 177 underground tanks in 18 separate tank farms containing accumulated liquid/sludge/salt cake radioactive wastes from 50 yr of weapons materials production activities. The total waste volume is about 60 million gal., which contains approximately 120 million Ci of radioactivity

  12. Probabilistic safety assessment for Hanford high-level waste tank 241-SY-101

    Energy Technology Data Exchange (ETDEWEB)

    MacFarlane, D.R.; Bott, T.F.; Brown, L.F.; Stack, D.W. [Los Alamos National Lab., NM (United States); Kindinger, J.; Deremer, R.K.; Medhekar, S.R.; Mikschl, T.J. [PLG, Inc., Newport Beach, CA (United States)

    1994-05-01

    Los Alamos National Laboratory (Los Alamos) is performing a comprehensive probabilistic safety assessment (PSA), which will include consideration of external events for the 18 tank farms at the Hanford Site. This effort is sponsored by the Department of Energy (DOE/EM, EM-36). Even though the methodology described herein will be applied to the entire tank farm, this report focuses only on the risk from the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases ({open_quotes}burps{close_quotes}) a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed first because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is being conducted in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. At the Hanford Site there are 177 underground tanks in 18 separate tank farms containing accumulated liquid/sludge/salt cake radioactive wastes from 50 yr of weapons materials production activities. The total waste volume is about 60 million gal., which contains approximately 120 million Ci of radioactivity.

  13. Assessment of chemical vulnerabilities in the Hanford high-level waste tanks

    International Nuclear Information System (INIS)

    Meacham, J.E.

    1996-01-01

    The purpose of this report is to summarize results of relevant data (tank farm and laboratory) and analysis related to potential chemical vulnerabilities of the Hanford Site waste tanks. Potential chemical safety vulnerabilities examined include spontaneous runaway reactions, condensed phase waste combustibility, and tank headspace flammability. The major conclusions of the report are the following: Spontaneous runaway reactions are not credible; condensed phase combustion is not likely; and periodic releases of flammable gas can be mitigated by interim stabilization

  14. Hanford high level waste (HLW) tank mixer pump safe operating envelope reliability assessment

    International Nuclear Information System (INIS)

    Fischer, S.R.; Clark, J.

    1993-01-01

    The US Department of Energy and its contractor, Westinghouse Corp., are responsible for the management and safe storage of waste accumulated from processing defense reactor irradiated fuels for plutonium recovery at the Hanford Site. These wastes, which consist of liquids and precipitated solids, are stored in underground storage tanks pending final disposition. Currently, 23 waste tanks have been placed on a safety watch list because of their potential for generating, storing, and periodically releasing various quantities of hydrogen and other gases. Tank 101-SY in the Hanford SY Tank Farm has been found to release hydrogen concentrations greater than the lower flammable limit (LFL) during periodic gas release events. In the unlikely event that an ignition source is present during a hydrogen release, a hydrogen burn could occur with a potential to release nuclear waste materials. To mitigate the periodic gas releases occurring from Tank 101-SY, a large mixer pump currently is being installed in the tank to promote a sustained release of hydrogen gas to the tank dome space. An extensive safety analysis (SA) effort was undertaken and documented to ensure the safe operation of the mixer pump after it is installed in Tank 101-SY.1 The SA identified a need for detailed operating, alarm, and abort limits to ensure that analyzed safety limits were not exceeded during pump operations

  15. Probability, consequences, and mitigation for lightning strikes to Hanford site high-level waste tanks

    Energy Technology Data Exchange (ETDEWEB)

    Zach, J.J.

    1996-08-01

    The purpose of this report is to summarize selected lightning issues concerning the Hanford Waste Tanks. These issues include the probability of lightning discharge striking the area immediately adjacent to a tank including a riser, the consequences of significant energy deposition from a lightning strike in a tank, and mitigating actions that have been or are being taken. The major conclusion of this report is that the probability of a lightning strike depositing sufficient energy in a tank to cause an effect on employees or the public is unlikely;but there are insufficient, quantitative data on the tanks and waste to prove that. Protection, such as grounding of risers and air terminals on existing light poles, is recommended.

  16. Sampling and analysis of high level waste tank supernatant: an overview

    International Nuclear Information System (INIS)

    Goergen, C.R.

    1981-01-01

    The Savannah River Plant routinely samples its high level radioactive waste tank supernatants for analysis of major components. These results are important in maintaining proper levels of corrosion inhibiters for protection of the tank walls. Because the tank ambient temperature is elevated, the sample is heated to 70 0 C prior to removing aliquots for use in a variety of analytical methods. Typical analyses include density, pH, OH - , NO 3 - , and NO 2 - , with occasional requests for Al(OH) 4 - , CO 3 /sup =/, PO 4 /sup =/, SO 4 /sup =/, and various radionuclides

  17. Derived Requirements for Double Shell Tank (DST) High Level Waste (HLW) Auxiliary Solids Mobilization

    Energy Technology Data Exchange (ETDEWEB)

    TEDESCHI, A.R.

    2000-02-28

    The potential need for auxiliary double-shell tank waste mixing and solids mobilization requires an evaluation of optional technologies. This document formalizes those operating and design requirements needed for further engineering evaluations.

  18. Evaluation of Flygt Propeller Xixers for Double Shell Tank (DST) High Level Waste Auxiliary Solids Mobilization

    Energy Technology Data Exchange (ETDEWEB)

    PACQUET, E.A.

    2000-07-20

    The River Protection Project (RPP) is planning to retrieve radioactive waste from the single-shell tanks (SST) and double-shell tanks (DST) underground at the Hanford Site. This waste will then be transferred to a waste treatment plant to be immobilized (vitrified) in a stable glass form. Over the years, the waste solids in many of the tanks have settled to form a layer of sludge at the bottom. The thickness of the sludge layer varies from tank to tank, from no sludge or a few inches of sludge to about 15 ft of sludge. The purpose of this technology and engineering case study is to evaluate the Flygt{trademark} submersible propeller mixer as a potential technology for auxiliary mobilization of DST HLW solids. Considering the usage and development to date by other sites in the development of this technology, this study also has the objective of expanding the knowledge base of the Flygt{trademark} mixer concept with the broader perspective of Hanford Site tank waste retrieval. More specifically, the objectives of this study delineated from the work plan are described.

  19. Savannah River Site High-Level Waste Tank Closure Final Environmental Impact Statement

    International Nuclear Information System (INIS)

    2002-01-01

    The U.S. Atomic Energy Commission, a U.S. Department of Energy (DOE) predecessor agency, established the Savannah River Site (SRS) near Aiken, South Carolina, in the early 1950s. The primary mission of SRS was to produce nuclear materials for national defense. With the end of the Cold War and the reduction in the size of the United States stockpile of nuclear weapons, the SRS mission has changed. While national defense is still an important facet of the mission, SRS no longer produces nuclear materials and the mission is focused on material stabilization, environmental restoration, waste management, and decontamination and decommissioning of facilities that are no longer needed. As a result of its nuclear materials production mission, SRS generated large quantities of high-level radioactive waste (HLW). The HLW resulted from dissolving spent reactor fuel and nuclear targets to recover the valuable radioactive isotopes. DOE had stored the HLW in 51 large underground storage tanks located in the F- and H-Area Tank Farms at SRS. DOE has emptied and closed two of those tanks. DOE is treating the HLW, using a process called vitrification. The highly radioactive portion of the waste is mixed with a glass like material and stored in stainless steel canisters at SRS, pending shipment to a geologic repository for disposal. This process is currently underway at SRS in the Defense Waste Processing Facility (DWPF). The HLW tanks at SRS are of four different types, which provide varying degrees of protection to the environment due to different degrees of containment. The tanks are operated under the authority of the Atomic Energy Act of 1954 (AEA) and DOE Orders issued under the AEA. The tanks are permitted by the South Carolina Department of Health and Environmental Control (SCDHEC) under South Carolina wastewater regulations, which require permitted facilities to be closed after they are removed from service. DOE has entered into an agreement with the U.S. Environmental

  20. Considerations of fluid-structure interaction effects in the design of high-level waste storage tanks

    International Nuclear Information System (INIS)

    Stuart, R.J.; Shipley, L.E.; Ghose, A.; Hiremath, M.S.

    1994-01-01

    For the seismic evaluation and design of the large number of underground high-level waste storage tanks (HLWST) at DOE sites, an important consideration is the adequate estimation of the fluid-structure interaction effects on the design forces. The DOE Tanks Seismic Experts Panel (TSEP) has developed seismic design and evaluation guidelines which include simplified methods for estimating hydrodynamic effects on tanks. For the practical analysis and design of HLWSTs, however, more sophisticated methods are often needed. The research presented in this paper demonstrates the effectiveness and reliability of finite element method based techniques, developed and utilized by ARES, to evaluate the fluid-structure interaction effects on underground HLWSTs. Analysis results for simple cylindrical tank configurations are first compared with previously published data, to benchmark the techniques. Next, for an actual HLWST configuration, correlations are established between these techniques and the TSEP guidelines, for the design parameters affected by fluid-structure interaction. Finally, practical design situations which may require a level of analysis sophistication that goes beyond the simplified TSEP guidelines are presented. This level of sophistication is frequently required when attempting to validate or upgrade the design qualifications of existing tanks

  1. Removal of salt from high-level waste tanks by density-driven circulation or mechanical agitation

    International Nuclear Information System (INIS)

    Kiser, D.L.

    1981-01-01

    Twenty-two high-level waste storage tanks at the Savannah River Plant are to be retired in the tank replacement/waste transfer program. The salt-removal portion of this program requires dissolution of about 19 million liters of salt cake. Steam circulation jets were originally proposed to dissolve the salt cake. However, the jets heated the waste tank to 80 to 90 0 C. This high temperature required a long cooldown period before transfer of the supernate by jet, and increased the risk of stress-corrosion cracking in these older tanks. A bench-scale investigation at the Savannah River Laboratory developed two alternatives to steam-jet circulation. One technique was density-driven circulation, which in bench tests dissolved salt at the same rate as a simulated steam circulation jet but at a lower temperature. The other technique was mechanical agitation, which dissolved the salt cake faster and required less fresh water than either density-driven circulation or the simulated steam circulation jet. Tests in an actual waste tank verified bench-scale results and demonstrated the superiority of mechanical agitation

  2. A dynamic simulation model of the Savannah River High Level Waste Tank Farm

    International Nuclear Information System (INIS)

    Gregory, M.V.; Shanahan, K.L.; Hang, T.

    1993-01-01

    A detailed, dynamic simulation model of a 51-tank high level radioactive waste storage and processing complex is being developed using SPEEDUP TM software. The initial model represents mass transfer, evaporation, precipitation, and sludge washing unit operation processes in about one-third of the tank farm complex through the solution of more than of 1000 coupled differential and algebraic equations, Eight discrete chemical constituents are tracked throughout the unit operations. The model is built upon FORTRAN both automatically generated by SPEEDUP TM to represent continuous processes, and custom-tailored by ourselves to control batched unit operations. The current model requires under 4 CPU seconds per simulated day on a VAX 8810. Strategies have been developed for expanding the model to the full tank farm complex

  3. Planning exercise for the resolution of high level waste tank safety issues

    International Nuclear Information System (INIS)

    Bunting, J.; Saveland, J.

    1992-01-01

    Several conditions have been found to exist within high level radioactive waste storage tanks at the Hanford site which could lead to uncontrolled exothermic reactions and/or to the release of tank contents into the environment. These conditions have led to the establishment of four priority 1 safety issues for the Hanford tanks. Resolution of these safety issues will require the coordinated efforts of professionals in chemical, nuclear, operations, safety, and other technical areas. A coordinated and integrated approach is necessary in order to achieve resolution in the shortest possible time, while ensuring that the steps taken do not complicate the later jobs of vitrification and ultimate disposal. This paper describes the purpose, process, and results of an effort to develop a suggested approach. (author)

  4. Chemical dissolving of sludge from a high level waste tank at the Savannah River Plant

    International Nuclear Information System (INIS)

    Bradley, R.F.; Hill, A.J. Jr.

    1977-11-01

    The concept for decontamination and retirement of radioactive liquid waste tanks at the Savannah River Plant (SRP) involves hydraulic slurrying to remove most of the settled sludges followed by chemical dissolving of residual sludges. Dissolving tests were carried out with small samples of sludge from SRP Tank 16H. Over 95 percent of the sludge was dissolved by 8 wt percent oxalic acid at 85 0 C with agitation in a two-step dissolving process (50 hours per step) and an initial reagent-to-sludge volume of 20. Oxalic acid does not attack the waste tank material of construction, appears to be compatible with the existing waste farm processes and equipment after neutralization, and with future processes planned for fixation of the waste into a high-integrity solid for packaging and shipping

  5. Nonliner analysis techniques for use in the assessment of high-level waste storage tank structures

    International Nuclear Information System (INIS)

    Moore, C.J.; Julyk, L.J.; Fox, G.L.; Dyrness, A.D.

    1991-09-01

    Reinforced concrete in combination with a steel liner has had a wide application to structures containing hazardous material. The buried double-shell waste storage tanks at the US Department of Energy's Hanford Site use this construction method. The generation and potential ignition of combustible gases within the primary tank is postulated to develop beyond-design-basis internal pressure and possible impact loading. The scope of this paper includes the illustration of analysis techniques for the assessment of these beyond-design-basis loadings. The analysis techniques include the coupling of the gas dynamics with the structural response, the treatment of reinforced concrete in regimes of inelastic behavior, and the treatment of geometric nonlinearities. The techniques and software tools presented provide a powerful nonlinear analysis capability for storage tanks. 10 refs., 13 figs., 1 tab

  6. Nonlinear analysis techniques for use in the assessment of high-level waste tank structures

    International Nuclear Information System (INIS)

    Moore, C.J.; Julyk, L.J.; Fox, G.L.; Dyrness, A.D.

    1991-01-01

    Reinforced concrete in combination with a steel liner has had a wide application to structures containing hazardous material. The buried double-shell waste storage tanks at the US Department of Energy's Hanford Site use this construction method. The generation and potential ignition of combustible gases within the primary tank is postulated to develop beyond-design-basis internal pressure and possible impact loading. The scope of this paper includes the illustration of analysis techniques for the assessment of these beyond-design-basis loadings. The analysis techniques include the coupling of the gas dynamics with the structural response, the treatment of reinforced concrete in regimes of inelastic behavior, and the treatment of geometric nonlinearities. The techniques and software tools presented provide a powerful nonlinear analysis capability for storage tanks

  7. Annual report, spring 2015. Alternative chemical cleaning methods for high level waste tanks-corrosion test results

    Energy Technology Data Exchange (ETDEWEB)

    Wyrwas, R. B. [Savannah River Site (SRS), Aiken, SC (United States)

    2015-07-06

    The testing presented in this report is in support of the investigation of the Alternative Chemical Cleaning program to aid in developing strategies and technologies to chemically clean radioactive High Level Waste tanks prior to tank closure. The data and conclusions presented here were the examination of the corrosion rates of A285 carbon steel and 304L stainless steel when interacted with the chemical cleaning solution composed of 0.18 M nitric acid and 0.5 wt. % oxalic acid. This solution has been proposed as a dissolution solution that would be used to remove the remaining hard heel portion of the sludge in the waste tanks. This solution was combined with the HM and PUREX simulated sludge with dilution ratios that represent the bulk oxalic cleaning process (20:1 ratio, acid solution to simulant) and the cumulative volume associated with multiple acid strikes (50:1 ratio). The testing was conducted over 28 days at 50°C and deployed two methods to invest the corrosion conditions; passive weight loss coupon and an active electrochemical probe were used to collect data on the corrosion rate and material performance. In addition to investigating the chemical cleaning solutions, electrochemical corrosion testing was performed on acidic and basic solutions containing sodium permanganate at room temperature to explore the corrosion impacts if these solutions were to be implemented to retrieve remaining actinides that are currently in the sludge of the tank.

  8. Seismic design and evaluation guidelines for the Department of Energy High-Level Waste Storage Tanks and Appurtenances

    International Nuclear Information System (INIS)

    Bandyopadhyay, K.; Cornell, A.; Costantino, C.; Kennedy, R.; Miller, C.; Veletsos, A.

    1995-10-01

    This document provides seismic design and evaluation guidelines for underground high-level waste storage tanks. The guidelines reflect the knowledge acquired in the last two decades in defining seismic ground motion and calculating hydrodynamic loads, dynamic soil pressures and other loads for underground tank structures, piping and equipment. The application of the guidelines is illustrated with examples. The guidelines are developed for a specific design of underground storage tanks, namely double-shell structures. However, the methodology discussed is applicable for other types of tank structures as well. The application of these and of suitably adjusted versions of these concepts to other structural types will be addressed in a future version of this document. The original version of this document was published in January 1993. Since then, additional studies have been performed in several areas and the results are included in this revision. Comments received from the users are also addressed. Fundamental concepts supporting the basic seismic criteria contained in the original version have since then been incorporated and published in DOE-STD-1020-94 and its technical basis documents. This information has been deleted in the current revision

  9. Seismic design and evaluation guidelines for the Department of Energy High-Level Waste Storage Tanks and Appurtenances

    Energy Technology Data Exchange (ETDEWEB)

    Bandyopadhyay, K.; Cornell, A.; Costantino, C.; Kennedy, R.; Miller, C.; Veletsos, A.

    1995-10-01

    This document provides seismic design and evaluation guidelines for underground high-level waste storage tanks. The guidelines reflect the knowledge acquired in the last two decades in defining seismic ground motion and calculating hydrodynamic loads, dynamic soil pressures and other loads for underground tank structures, piping and equipment. The application of the guidelines is illustrated with examples. The guidelines are developed for a specific design of underground storage tanks, namely double-shell structures. However, the methodology discussed is applicable for other types of tank structures as well. The application of these and of suitably adjusted versions of these concepts to other structural types will be addressed in a future version of this document. The original version of this document was published in January 1993. Since then, additional studies have been performed in several areas and the results are included in this revision. Comments received from the users are also addressed. Fundamental concepts supporting the basic seismic criteria contained in the original version have since then been incorporated and published in DOE-STD-1020-94 and its technical basis documents. This information has been deleted in the current revision.

  10. High Level Waste Tank Farm Replacement Project for the Idaho Chemical Processing Plant at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    1993-06-01

    The Department of Energy (DOE) has prepared an environmental assessment (EA), DOE/EA-0831, for the construction and operation of the High-Level Waste Tank Farm Replacement (HLWTFR) Project for the Idaho Chemical Processing Plant located at the Idaho National Engineering Laboratory (INEL). The HLWTFR Project as originally proposed by the DOE and as analyzed in this EA included: (1) replacement of five high-level liquid waste storage tanks with four new tanks and (2) the upgrading of existing tank relief piping and high-level liquid waste transfer systems. As a result of the April 1992 decision to discontinue the reprocessing of spent nuclear fuel at INEL, DOE believes that it is unlikely that the tank replacement aspect of the project will be needed in the near term. Therefore, DOE is not proposing to proceed with the replacement of the tanks as described in this-EA. The DOE's instant decision involves only the proposed upgrades aspect of the project described in this EA. The upgrades are needed to comply with Resource Conservation and Recovery Act, the Idaho Hazardous Waste Management Act requirements, and the Department's obligations pursuant to the Federal Facilities Compliance Agreement and Consent Order among the Environmental Protection Agency, DOE, and the State of Idaho. The environmental impacts of the proposed upgrades are adequately covered and are bounded by the analysis in this EA. If DOE later proposes to proceed with the tank replacement aspect of the project as described in the EA or as modified, it will undertake appropriate further review pursuant to the National Environmental Policy Act

  11. High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 7

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    This Requirements Identification Document (RID) describes an Occupational Health and Safety Program as defined through the Relevant DOE Orders, regulations, industry codes/standards, industry guidance documents and, as appropriate, good industry practice. The definition of an Occupational Health and Safety Program as specified by this document is intended to address Defense Nuclear Facilities Safety Board Recommendations 90-2 and 91-1, which call for the strengthening of DOE complex activities through the identification and application of relevant standards which supplement or exceed requirements mandated by DOE Orders. This RID applies to the activities, personnel, structures, systems, components, and programs involved in maintaining the facility and executing the mission of the High-Level Waste Storage Tank Farms.

  12. High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 7

    International Nuclear Information System (INIS)

    1994-04-01

    This Requirements Identification Document (RID) describes an Occupational Health and Safety Program as defined through the Relevant DOE Orders, regulations, industry codes/standards, industry guidance documents and, as appropriate, good industry practice. The definition of an Occupational Health and Safety Program as specified by this document is intended to address Defense Nuclear Facilities Safety Board Recommendations 90-2 and 91-1, which call for the strengthening of DOE complex activities through the identification and application of relevant standards which supplement or exceed requirements mandated by DOE Orders. This RID applies to the activities, personnel, structures, systems, components, and programs involved in maintaining the facility and executing the mission of the High-Level Waste Storage Tank Farms

  13. High-level waste tank remediation technology integration summary. Revision 1

    International Nuclear Information System (INIS)

    DeLannoy, C.R.; Susiene, C.; Fowler, K.M.; Robson, W.M.; Cruse, J.M.

    1994-07-01

    The U.S. Department of Energy's Environmental Restoration and Waste Management and Technology Development Programs are engaged in a number of projects to develop, demonstrate, test, and evaluate new technologies to support the cleanup and site remediation of more than 300 underground storage tanks containing over 381,000 m 3 (100 million gal) of liquid radioactive mixed waste at the Hanford Reservation. Significant development is needed within primary functions and in determining an overall bounding strategy. This document is an update of continuing work to summarize the overall strategy and to provide data regarding technology development activities within the strategy. It is intended to serve as an information resource to support understanding, decision making, and integration of multiple program technology development activities. Recipients are encouraged to provide comments and input to the authors for incorporation in future revisions

  14. The Apparent Solubility Of Aluminum(III) In Hanford High-Level Waste Tanks

    International Nuclear Information System (INIS)

    Reynolds, J.G.

    2012-01-01

    The solubility of aluminum in Hanford nuclear waste impacts on the process ability of the waste by a number of proposed treatment options. For many years, Hanford staff has anecdotally noted that aluminum appears to be considerably more soluble in Hanford waste than the simpler electrolyte solutions used as analogues. There has been minimal scientific study to confirm these anecdotal observations, however. The present study determines the apparent solubility product for gibbsite in 50 tank samples. The ratio of hydroxide to aluminum in the liquid phase for the samples is calculated and plotted as a function of total sodium molarity. Total sodium molarity is used as a surrogate for ionic strength, because the relative ratios of mono, di and trivalent anions are not available for all of the samples. These results were compared to the simple NaOH-NaAl(OH 4 )H 2 O system, and the NaOH-NaAl(OH 4 )NaCl-H 2 O system data retrieved from the literature. The results show that gibbsite is apparently more soluble in the samples than in the simple systems whenever the sodium molarity is greater than two. This apparent enhanced solubility cannot be explained solely by differences in ionic strength. The change in solubility with ionic strength in simple systems is small compared to the difference between aluminum solubility in Hanford waste and the simple systems. The reason for the apparent enhanced solubility is unknown, but could include. kinetic or thermodynamic factors that are not present in the simple electrolyte systems. Any kinetic explanation would have to explain why the samples are always supersaturated whenever the sodium molarity is above two. Real waste characterization data should not be used to validate thermodynamic solubility models until it can be confirmed that the apparent enhanced gibbsite solubility is a thermodynamic effect and not a kinetic effect.

  15. High Performance Zero-Bleed CLSM/Grout Mixes for High-Level Waste Tank Closures Strategic Research and Development - FY99 Report

    International Nuclear Information System (INIS)

    Langton, C.A.

    2000-01-01

    The overall objective of this program, SRD-99-08, was to design and test suitable materials, which can be used to close high-level waste tanks at SRS. Fill materials can be designed to perform several functions including chemical stabilization and/or physical encapsulation of incidental waste so that the potential for transport of contaminants into the environment is reduced. Also they are needed to physically stabilize the void volume in the tanks to prevent/minimize future subsidence and inadvertent intrusion. The intent of this work was to develop a zero-bleed soil CLSM (ZBS-CLSM) and a zero-bleed concrete mix (ZBC) which meet the unique placement and stabilization/encapsulation requirements for high-level waste tank closures. These mixes in addition to the zero-bleed CLSM mixes formulated for closure of Tanks 17-F and 20-F provide design engineers with a suite of options for specifying materials for future tank closures

  16. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 7. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Burt, D.L.

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 7) presents the standards and requirements for the following sections: Occupational Safety and Health, and Environmental Protection.

  17. High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 7. Revision 1

    International Nuclear Information System (INIS)

    Burt, D.L.

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 7) presents the standards and requirements for the following sections: Occupational Safety and Health, and Environmental Protection

  18. Status of the Development of In-Tank/At-Tank Separations Technologies for High-Level Waste Processing for the U.S. Department of Energy

    Energy Technology Data Exchange (ETDEWEB)

    Wilmarth, William R.; Machara, N.; Peterson, Reid A.; Bush, Sheryl R.

    2011-10-01

    Within the U.S. Department of Energy’s (DOE) Office of Technology Innovation and Development, the Office of Waste Processing manages a research and development program related to the treatment and disposition of radioactive waste. At the Savannah River (South Carolina) and Hanford (Washington) Sites, approximately 90 million gallons of waste are distributed among 226 storage tanks (grouped or collocated in “tank farms”). This waste may be considered to contain mixed and stratified high activity and low activity constituent waste liquids, salts and sludges that are collectively managed as high level waste (HLW). A large majority of these wastes and associated facilities are unique to the DOE, meaning many of the programs to treat these materials are “first-of-a-kind” and unprecedented in scope and complexity. As a result, the technologies required to disposition these wastes must be developed from basic principles, or require significant re-engineering to adapt to DOE’s specific applications. Of particular interest recently, the development of In-tank or At-Tank separation processes have the potential to treat waste with high returns on financial investment. The primary objective associated with In-Tank or At-Tank separation processes is to accelerate waste processing. Insertion of the technologies will (1) maximize available tank space to efficiently support permanent waste disposition including vitrification; (2) treat problematic waste prior to transfer to the primary processing facilities at either site (i.e., Hanford’s Waste Treatment and Immobilization Plant (WTP) or Savannah River’s Salt Waste Processing Facility (SWPF)); and (3) create a parallel treatment process to shorten the overall treatment duration.

  19. High performance zero-bleed CLSM/grout mixes for high-level waste tank closures strategic research and development - FY98

    International Nuclear Information System (INIS)

    Langton, C.A.

    2000-01-01

    The overall objective of this program, SRD-98-08, is to design and test suitable materials, which can be used to close high-level waste tanks at the Savannah River Site. Fill materials can be designed to perform several functions. They can be designed to chemically stabilize and/or physically encapsulate incidental waste so that the potential for transport of contaminants into the environment is reduced. Also they are needed to physically stabilize the void volume in the tanks to prevent/minimize future subsidence and inadvertent intrusion

  20. Status of the development of in-tank/at-tank separations technologies for high-level waste processing for the U.S. Department Of Energy - 59109

    International Nuclear Information System (INIS)

    Wilmarth, William R.; Bush, Sheryl R.; Machara, Nicholas P.; Peterson, Reid A.

    2012-01-01

    Within the U.S. Department of Energy's (DOE) Office of Technology Innovation and Development, the Office of Waste Processing manages a research and development program related to the treatment and disposition of radioactive waste. At the Savannah River (South Carolina) and Hanford (Washington) Sites, approximately 90 million gallons of waste are distributed among 226 storage tanks (grouped or collocated in 'tank farms'). This waste may be considered to contain mixed and stratified high activity and low activity constituent waste liquids, salts and sludges that are collectively managed as high level waste (HLW). A large majority of these wastes and associated facilities are unique to the DOE, meaning many of the programs to treat these materials are 'first-of-a-kind' and unprecedented in scope and complexity. As a result, the technologies required to disposition these wastes must be developed from basic principles, or require significant re-engineering to adapt to DOE's specific applications. Of particular interest recently, the development of In-tank or At- Tank separation processes have the potential to treat waste with high returns on financial investment. The primary objective associated with In-Tank or At-Tank separation processes is to accelerate waste processing. Insertion of the technologies will (1) maximize available tank space to efficiently support permanent waste disposition including vitrification; (2) treat problematic waste prior to transfer to the primary processing facilities at either site (i.e., Hanford's Waste Treatment and Immobilization Plant (WTP) or Savannah River's Salt Waste Processing Facility (SWPF)); and (3) create a parallel treatment process to shorten the overall treatment duration. This paper will review the status of several of the R and D projects being developed by the U.S. DOE including insertion of the ion exchange (IX) technologies, such as Small Column Ion Exchange (SCIX) at Savannah River. This has the potential to align the

  1. STATUS OF THE DEVELOPMENT OF IN-TANK/AT-TANK SEPARATIONS TECHNOLOGIES FOR FOR HIGH-LEVEL WASTE PROCESSING FOR THE U.S. DEPARTMENT OF ENERGY

    Energy Technology Data Exchange (ETDEWEB)

    Aaron, G.; Wilmarth, B.

    2011-09-19

    Within the U.S. Department of Energy's (DOE) Office of Technology Innovation and Development, the Office of Waste Processing manages a research and development program related to the treatment and disposition of radioactive waste. At the Savannah River (South Carolina) and Hanford (Washington) Sites, approximately 90 million gallons of waste are distributed among 226 storage tanks (grouped or collocated in 'tank farms'). This waste may be considered to contain mixed and stratified high activity and low activity constituent waste liquids, salts and sludges that are collectively managed as high level waste (HLW). A large majority of these wastes and associated facilities are unique to the DOE, meaning many of the programs to treat these materials are 'first-of-a-kind' and unprecedented in scope and complexity. As a result, the technologies required to disposition these wastes must be developed from basic principles, or require significant re-engineering to adapt to DOE's specific applications. Of particular interest recently, the development of In-tank or At-Tank separation processes have the potential to treat waste with high returns on financial investment. The primary objective associated with In-Tank or At-Tank separation processes is to accelerate waste processing. Insertion of the technologies will (1) maximize available tank space to efficiently support permanent waste disposition including vitrification; (2) treat problematic waste prior to transfer to the primary processing facilities at either site (i.e., Hanford's Waste Treatment and Immobilization Plant (WTP) or Savannah River's Salt Waste Processing Facility (SWPF)); and (3) create a parallel treatment process to shorten the overall treatment duration. This paper will review the status of several of the R&D projects being developed by the U.S. DOE including insertion of the ion exchange (IX) technologies, such as Small Column Ion Exchange (SCIX) at Savannah River

  2. Status Of The Development Of In-Tank/At-Tank Separations Technologies For High-Level Waste Processing For The U.S. Department Of Energy

    International Nuclear Information System (INIS)

    Aaron, G.; Wilmarth, B.

    2011-01-01

    Within the U.S. Department of Energy's (DOE) Office of Technology Innovation and Development, the Office of Waste Processing manages a research and development program related to the treatment and disposition of radioactive waste. At the Savannah River (South Carolina) and Hanford (Washington) Sites, approximately 90 million gallons of waste are distributed among 226 storage tanks (grouped or collocated in 'tank farms'). This waste may be considered to contain mixed and stratified high activity and low activity constituent waste liquids, salts and sludges that are collectively managed as high level waste (HLW). A large majority of these wastes and associated facilities are unique to the DOE, meaning many of the programs to treat these materials are 'first-of-a-kind' and unprecedented in scope and complexity. As a result, the technologies required to disposition these wastes must be developed from basic principles, or require significant re-engineering to adapt to DOE's specific applications. Of particular interest recently, the development of In-tank or At-Tank separation processes have the potential to treat waste with high returns on financial investment. The primary objective associated with In-Tank or At-Tank separation processes is to accelerate waste processing. Insertion of the technologies will (1) maximize available tank space to efficiently support permanent waste disposition including vitrification; (2) treat problematic waste prior to transfer to the primary processing facilities at either site (i.e., Hanford's Waste Treatment and Immobilization Plant (WTP) or Savannah River's Salt Waste Processing Facility (SWPF)); and (3) create a parallel treatment process to shorten the overall treatment duration. This paper will review the status of several of the R and D projects being developed by the U.S. DOE including insertion of the ion exchange (IX) technologies, such as Small Column Ion Exchange (SCIX) at Savannah River. This has the potential to align the

  3. Status of containment integrity studies for continued in-tank storage of Hanford defense high-level waste

    International Nuclear Information System (INIS)

    Baca, R.G.; Beitel, G.A.; Mercier, P.F.; Moore, E.L.; Vollert, F.R.

    1978-09-01

    Information is provided on the technical studies that have been implemented for evaluating the containment integrity of the single-shell waste storage tanks. The major areas of study are an analysis of storage tank integrity, a failure mode analysis, and storage tank improvements. Evaluations of tank structural integrity include theoretical studies on static and dynamic load responses, laboratory studies on concrete durability, and experimental studies on the potential for exothermic reactions of salt cake. The structural analyses completed to date show that the tanks are in good condition and have a safety margin against overload. Environmental conditions that could cause a loss of durability are limited to the waste chemicals stored (which do not have access to the concrete). Concern that a salt cake exothermic reaction may initiate a loss of containment is not justifiable based on extensive testing completed. A failure mode analysis of a tank liner failure, a sidewall failure, and a dome collapse shows that no radiologic hazard to man results. Storage tank improvement studies completed show that support of a tank dome is achievable. Secondary containment provided by chemical grouts and bentonite clay slurry walls does not appear promising. It is now estimated that the single-shell tanks will be serviceable for the storage of salt cake waste for decades under currently established operating temperature and load limits

  4. Risk-based systems analysis of emerging high-level waste tank remediation technologies. Volume 2: Final report

    International Nuclear Information System (INIS)

    Peters, B.B.; Cameron, R.J.; McCormack, W.D.

    1994-08-01

    The objective of DOE's Radioactive Waste Tank Remediation Technology Focus Area is to identify and develop new technologies that will reduce the risk and/or cost of remediating DOE underground waste storage tanks and tank contents. There are, however, many more technology investment opportunities than the current budget can support. Current technology development selection methods evaluate new technologies in isolation from other components of an overall tank waste remediation system. This report describes a System Analysis Model developed under the US Department of Energy (DOE) Office of Technology Development (OTD) Underground Storage Tank-Integrated Demonstration (UST-ID) program. The report identifies the project objectives and provides a description of the model. Development of the first ''demonstration'' version of this model and a trial application have been completed and the results are presented. This model will continue to evolve as it undergoes additional user review and testing

  5. Environmental Assessment for the Closure of the High-Level Waste Tanks in F- and H-Areas at the Savannah River Site

    International Nuclear Information System (INIS)

    1996-01-01

    This Environmental Assessment (EA) has been prepared by the Department of Energy (DOE) to assess the potential environmental impacts associated with the closure of 51 high-level radioactive waste tanks and tank farm ancillary equipment (including transfer lines, evaporators, filters, pumps, etc) at the Savannah River Site (SRS) located near Aiken, South Carolina. The waste tanks are located in the F- and H-Areas of SRS and vary in capacity from 2,839,059 liters (750,000 gallons) to 4,921,035 liters (1,300,000 gallons). These in-ground tanks are surrounded by soil to provide shielding. The F- and H-Area High-Level Waste Tanks are operated under the authority of Industrial Wastewater Permits No.17,424-IW; No.14520, and No.14338 issued by the South Carolina Department of Health and Environmental Control (SCDHEC). In accordance with the Permit requirements, DOE has prepared a Closure Plan (DOE, 1996) and submitted it to SCDHEC for approval. The Closure Plan identifies all applicable or relevant and appropriate regulations, statutes, and DOE Orders for closing systems operated under the Industrial Wastewater Permits. When approved by SCDHEC, the Closure Plan will present the regulatory process for closing all of the F- and H-Area High Level Waste Tanks. The Closure Plan establishes performance objectives or criteria to be met prior to closing any tank, group of tanks, or ancillary tank farm equipment. The proposed action is to remove the residual wastes from the tanks and to fill the tanks with a material to prevent future collapse and bind up residual waste, to lower human health risks, and to increase safety in and around the tanks. If required, an engineered cap consisting of clay, backfill (soil), and vegetation as the final layer to prevent erosion would be applied over the tanks. The selection of tank system closure method will be evaluated against the following Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) criteria described in 40

  6. High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 4

    International Nuclear Information System (INIS)

    1994-04-01

    Radiation protection of personnel and the public is accomplished by establishing a well defined Radiation Protection Organization to ensure that appropriate controls on radioactive materials and radiation sources are implemented and documented. This Requirements Identification Document (RID) applies to the activities, personnel, structures, systems, components, and programs involved in executing the mission of the Tank Farms. The physical boundaries within which the requirements of this RID apply are the Single Shell Tank Farms, Double Shell Tank Farms, 242-A Evaporator-Crystallizer, 242-S, T Evaporators, Liquid Effluent Retention Facility (LERF), Purgewater Storage Facility (PWSF), and all interconnecting piping, valves, instrumentation, and controls. Also included is all piping, valves, instrumentation, and controls up to and including the most remote valve under Tank Farms control at any other Hanford Facility having an interconnection with Tank Farms. The boundary of the structures, systems, components, and programs to which this RID applies, is defined by those that are dedicated to and/or under the control of the Tank Farms Operations Department and are specifically implemented at the Tank Farms

  7. High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 4

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    Radiation protection of personnel and the public is accomplished by establishing a well defined Radiation Protection Organization to ensure that appropriate controls on radioactive materials and radiation sources are implemented and documented. This Requirements Identification Document (RID) applies to the activities, personnel, structures, systems, components, and programs involved in executing the mission of the Tank Farms. The physical boundaries within which the requirements of this RID apply are the Single Shell Tank Farms, Double Shell Tank Farms, 242-A Evaporator-Crystallizer, 242-S, T Evaporators, Liquid Effluent Retention Facility (LERF), Purgewater Storage Facility (PWSF), and all interconnecting piping, valves, instrumentation, and controls. Also included is all piping, valves, instrumentation, and controls up to and including the most remote valve under Tank Farms control at any other Hanford Facility having an interconnection with Tank Farms. The boundary of the structures, systems, components, and programs to which this RID applies, is defined by those that are dedicated to and/or under the control of the Tank Farms Operations Department and are specifically implemented at the Tank Farms.

  8. Evaluation of Flygt Propeller Mixers for Double-Shell Tank (DST) High Level Waste Auxiliary Solids Mobilization

    International Nuclear Information System (INIS)

    PACQUET, E.A.

    2000-01-01

    The River Protection Project (RPP) is planning to retrieve radioactive waste from the single-shell tanks (SST) and double-shell tanks (DST) underground at the Hanford Site. This waste will then be transferred to a waste treatment plant to be immobilized (vitrified) in a stable glass form. Over the years, the waste solids in many of the tanks have settled to form a layer of sludge at the bottom. The thickness of the sludge layer varies from tank to tank, from no sludge or a few inches of sludge to about 15 ft of sludge. The purpose of this technology and engineering case study is to evaluate the Flygt(trademark) submersible propeller mixer as a potential technology for auxiliary mobilization of DST HLW solids. Considering the usage and development to date by other sites in the development of this technology, this study also has the objective of expanding the knowledge base of the Flygt(trademark) mixer concept with the broader perspective of Hanford Site tank waste retrieval. More specifically, the objectives of this study delineated from the work plan are described

  9. A Value of Information approach to data quality objectives for the Hanford high-level waste tanks

    International Nuclear Information System (INIS)

    Wood, T.W.; Hunter, V.L.; Ulvila, J.W.

    1995-02-01

    This report summarizes a Pacific Northwest Laboratory review of the organic-nitrate reaction safety issue in the Hanford single-shell tanks. This study employed a decision analytic method known as Value of Information (VOI). VOI analysis is a special form of decision analysis that has an information collection alternative as one of the initial decision choices. This type of decision analysis, therefore results in the ability to specify the preferred information collection alternative, taking into account all information gathering and other relevant alternatives. For example, the risk reduction benefit associated with further sampling to quantify total organic carbon inventory or to improve information on energetics can be compared to the risk reduction benefit of better temperature monitoring, operational restrictions, or mitigation by moisture control. This approach allows freedom from built-in assumptions, e.g., that all tanks must be sampled to some degree or that all tanks must be deemed intrinsically safe by some means or another. It allows for each tank management decision to be judged in terms of risk reduction from the current state of affairs, and for that state of affairs to be continuously updated to incorporate new information on tank contents, the phenomenology of safety issues, or the effectiveness of mitigation schemes

  10. High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 3

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The Safeguards and Security (S&S) Functional Area address the programmatic and technical requirements, controls, and standards which assure compliance with applicable S&S laws and regulations. Numerous S&S responsibilities are performed on behalf of the Tank Farm Facility by site level organizations. Certain other responsibilities are shared, and the remainder are the sole responsibility of the Tank Farm Facility. This Requirements Identification Document describes a complete functional Safeguards and Security Program that is presumed to be the responsibility of the Tank Farm Facility. The following list identifies the programmatic elements in the S&S Functional Area: Program Management, Protection Program Scope and Evaluation, Personnel Security, Physical Security Systems, Protection Program Operations, Material Control and Accountability, Information Security, and Key Program Interfaces.

  11. High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 5

    International Nuclear Information System (INIS)

    1994-04-01

    The Fire Protection functional area for the Hanford Site Tank Farm facilities and support structures is based on the application of relevant DOE orders, regulations, and industry codes and standards. The fire protection program defined in this document may be divided into three areas: (1) organizational, (2) administrative programmatic features, and (3) technical features. The information presented in each section is in the form of program elements and orders, regulations, industry codes, and standards that serve as the attributes of a fire protection program for the Tank Farm facilities. Upon completion this document will be utilized as the basis to evaluate compliance of the fire protection program being implemented for the Tank Farm facilities with the requirements of DOE orders and industry codes and standards

  12. Chemical speciation of strontium, americium, and curium in high level waste: Predictive modeling of phase partitioning during tank processing. Annual progress report, October 1996--September 1997

    International Nuclear Information System (INIS)

    Choppin, G.; Felmy, A.R.

    1997-01-01

    'The program at Florida State University was funded to collaborate with Dr. A. Felmy (PNNL) on speciation in high level wastes and with Dr. D. Rai (PNNL) on redox of Pu under high level waste conditions. The funding provided support for 3 research associates (postdoctoral researchers) under Professor G. R. Choppin as P.I. Dr. Kath Morris from U. Manchester (Great Britain), Dr. Dean Peterman and Dr. Amy Irwin (both from U. Cincinnati) joined the laboratory in the latter part of 1996. After an initial training period to become familiar with basic actinide chemistry and radiochemical techniques, they began their research. Dr. Peterman was assigned the task of measuring Th-EDTA complexation prior to measuring Pu(IV)-EDTA complexation. These studies are associated with the speciation program with Dr. Felmy. Drs. Morris and Irwin initiated research on redox of plutonium with agents present in the Hanford Tanks as a result of radiolysis or from use in separations. The preliminary results obtained thus far are described in this report. It is expected that the rate of progress will continue to increase significantly as the researchers gain more experience with plutonium chemistry.'

  13. Risk-based systems analysis of emerging high-level waste tank remediation technologies. Volume 1: Executive summary

    International Nuclear Information System (INIS)

    Peters, B.B.; Cameron, R.J.; McCormack, W.D.

    1994-08-01

    This report describes a System Analysis Model developed under the US Department of Energy (DOE) Office of Technology Development (OTD) Underground Storage Tank-Integrated Demonstration (UST-ID) program to aid technology development funding decisions for radioactive tank waste remediation. Current technology development selection methods evaluate new technologies in isolation from other components of an overall tank waste remediation system. These methods do not show the relative effect of new technologies on tank remediation systems as a whole. Consequently, DOE may spend its resources on technologies that promise to improve a single function but have a small or possibly negative, impact on the overall system, or DOE may overlook a technology that does not address a high priority problem in the system but that does, if implemented, offer sufficient overall improvements. Systems engineering and detailed analyses often conducted under the National Environmental Policy Act (NEPA 1969) use a ''whole system'' approach but are costly, too time-consuming, and often not sufficiently focused to support the needs of the technology program decision-makers. An alternative approach is required to evaluate these systems impacts but still meet the budget and schedule needs of the technology program

  14. High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 6

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The scope of the Environmental Restoration and Waste Management (EM) Functional Area includes the programmatic controls associated with the management and operation of the Hanford Tank Farm Facility. The driving management organization implementing the programmatic controls is the Tank Farms Waste Management (WM)organization whose responsibilities are to ensure that performance objectives are established; and that measurable criteria for attaining objectives are defined and reflected in programs, policies and procedures. Objectives for the WM Program include waste minimization, establishment of effective waste segregation methods, waste treatment technology development, radioactive (low-level, high-level) hazardous and mixed waste transfer, treatment, and storage, applicability of a corrective action program, and management and applicability of a decontamination and decommissioning (D&D) program in future years.

  15. High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 6

    International Nuclear Information System (INIS)

    1994-04-01

    The scope of the Environmental Restoration and Waste Management (EM) Functional Area includes the programmatic controls associated with the management and operation of the Hanford Tank Farm Facility. The driving management organization implementing the programmatic controls is the Tank Farms Waste Management (WM)organization whose responsibilities are to ensure that performance objectives are established; and that measurable criteria for attaining objectives are defined and reflected in programs, policies and procedures. Objectives for the WM Program include waste minimization, establishment of effective waste segregation methods, waste treatment technology development, radioactive (low-level, high-level) hazardous and mixed waste transfer, treatment, and storage, applicability of a corrective action program, and management and applicability of a decontamination and decommissioning (D ampersand D) program in future years

  16. High Level Waste Tank Farm Replacement Project for the Idaho Chemical Processing Plant at the Idaho National Engineering Laboratory. Environmental Assessment

    Energy Technology Data Exchange (ETDEWEB)

    1993-06-01

    The Department of Energy (DOE) has prepared an environmental assessment (EA), DOE/EA-0831, for the construction and operation of the High-Level Waste Tank Farm Replacement (HLWTFR) Project for the Idaho Chemical Processing Plant located at the Idaho National Engineering Laboratory (INEL). The HLWTFR Project as originally proposed by the DOE and as analyzed in this EA included: (1) replacement of five high-level liquid waste storage tanks with four new tanks and (2) the upgrading of existing tank relief piping and high-level liquid waste transfer systems. As a result of the April 1992 decision to discontinue the reprocessing of spent nuclear fuel at INEL, DOE believes that it is unlikely that the tank replacement aspect of the project will be needed in the near term. Therefore, DOE is not proposing to proceed with the replacement of the tanks as described in this-EA. The DOE`s instant decision involves only the proposed upgrades aspect of the project described in this EA. The upgrades are needed to comply with Resource Conservation and Recovery Act, the Idaho Hazardous Waste Management Act requirements, and the Department`s obligations pursuant to the Federal Facilities Compliance Agreement and Consent Order among the Environmental Protection Agency, DOE, and the State of Idaho. The environmental impacts of the proposed upgrades are adequately covered and are bounded by the analysis in this EA. If DOE later proposes to proceed with the tank replacement aspect of the project as described in the EA or as modified, it will undertake appropriate further review pursuant to the National Environmental Policy Act.

  17. Characterization Of Supernate Samples From High Level Waste Tanks 13H, 30H, 37H, 39H, 45F, 46F and 49H

    Energy Technology Data Exchange (ETDEWEB)

    Stallings, M. E.; Barnes, M. J.; Peters, T. B.; Diprete, D. P.; Hobbs, D. T.; Fink, S. D.

    2005-06-15

    This document presents work conducted in support of technical needs expressed, in part, by the Engineering, Procurement, and Construction Contractor for the Salt Waste Processing Facility (SWPF). The Department of Energy (DOE) requested that Savannah River National Laboratory (SRNL) analyze and characterize supernate waste from seven selected High Level Waste (HLW) tanks to allow: classification of feed to be sent to the SWPF; verification that SWPF processes will be able to meet Saltstone Waste Acceptance Criteria (WAC); and updating of the Waste Characterization System (WCS) database. This document provides characterization data of samples obtained from Tanks 13H, 30H, 37H, 39H, 45F, 46F, and 49H and discusses results. Characterization of the waste tank samples involved several treatments and analysis at various stages of sample processing. These analytical stages included as-received liquid, post-dilution to 6.44 M sodium (target), post-acid digestion, post-filtration (at 3 filtration pore sizes), and after cesium removal using ammonium molybdophosphate (AMP). All tanks will require cesium removal as well as treatment with Monosodium Titanate (MST) for {sup 90}Sr (Strontium) decontamination. A small filtration effect for 90Sr was observed for six of the seven tank wastes. No filtration effects were observed for Pu (Plutonium), Np (Neptunium), U (Uranium), or Tc (Technetium); {sup 137}Cs (Cesium) concentration is ~E+09 pCi/mL for all the tank wastes. Tank 37H is significantly higher in {sup 90}Sr than the other six tanks. {sup 237}Np in the F-area tanks (45F and 46F) are at least 1 order of magnitude less than the H-Area tank wastes. The data indicate a constant ratio of {sup 99}Tc to Cs in the seven tank wastes. This indicates the Tc remains largely soluble in Savannah River Site (SRS) waste and partitions similarly with Cs. {sup 241}Am (Americium) concentration was low in the seven tank wastes. The majority of data were detection limit values, the largest being

  18. High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 1

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The purpose of this Requirements Identification Document (RID) section is to identify, in one location, all of the facility specific requirements and good industry practices which are necessary or important to establish an effective Issues Management Program for the Tank Farm Facility. The Management Systems Functional Area includes the site management commitment to environmental safety and health (ES&H) policies and controls, to compliance management, to development and management of policy and procedures, to occurrence reporting and corrective actions, resource and issue management, and to the self-assessment process.

  19. High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 1

    International Nuclear Information System (INIS)

    1994-04-01

    The purpose of this Requirements Identification Document (RID) section is to identify, in one location, all of the facility specific requirements and good industry practices which are necessary or important to establish an effective Issues Management Program for the Tank Farm Facility. The Management Systems Functional Area includes the site management commitment to environmental safety and health (ES ampersand H) policies and controls, to compliance management, to development and management of policy and procedures, to occurrence reporting and corrective actions, resource and issue management, and to the self-assessment process

  20. High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 2

    International Nuclear Information System (INIS)

    1994-04-01

    The Quality Assurance Functional Area Requirements Identification Document (RID), addresses the programmatic requirements that ensure risks and environmental impacts are minimized, ensure safety, reliability, and performance are maximized through the application of effective management systems commensurate with the risks posed by the Tank Farm Facility and its operation. This RID incorporates guidance intended to provide Tank Farms management with the necessary requirements information to develop, upgrade, or assess the effectiveness of a Quality Assurance Program in the performance of organizational and functional activities. Quality Assurance is defined as all those planned and systematic actions necessary to provide adequate confidence that a facility, structure, system, or component will perform satisfactorily and safely in service. This document will provide the specific requirements to meet DNFSB recommendations and the guidance provided in DOE Order 5700.6C, utilizing industry codes, standards, regulatory guidelines, and industry good practices that have proven to be essential elements for an effective and efficient Quality Assurance Program as the nuclear industry has matured over the last thirty years

  1. High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 2

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The Quality Assurance Functional Area Requirements Identification Document (RID), addresses the programmatic requirements that ensure risks and environmental impacts are minimized, ensure safety, reliability, and performance are maximized through the application of effective management systems commensurate with the risks posed by the Tank Farm Facility and its operation. This RID incorporates guidance intended to provide Tank Farms management with the necessary requirements information to develop, upgrade, or assess the effectiveness of a Quality Assurance Program in the performance of organizational and functional activities. Quality Assurance is defined as all those planned and systematic actions necessary to provide adequate confidence that a facility, structure, system, or component will perform satisfactorily and safely in service. This document will provide the specific requirements to meet DNFSB recommendations and the guidance provided in DOE Order 5700.6C, utilizing industry codes, standards, regulatory guidelines, and industry good practices that have proven to be essential elements for an effective and efficient Quality Assurance Program as the nuclear industry has matured over the last thirty years.

  2. WVNS Tank Farm Process Support: Experimental evaluation of an inert gas (nitrogen) to mitigate external corrosion of high-level waste storage tanks

    International Nuclear Information System (INIS)

    Elmore, M.R.

    1996-02-01

    Corrosion of the carbon steel waste storage tanks at West Valley Nuclear Services continues to be of concern, especially as the planned duration of waste storage time increases and sludge washing operations are conducted. The external surfaces of Tanks 8D-1 and 8D-2 have been exposed for more than 10 years to water that has intruded into the tank vaults. Visual inspection of the external tank surfaces using a remote video camera has shown indications of heavy corrosion in localized areas on the tank walls. Tests on mild steel specimens under simulated tank vault conditions showed that corrosion is related to the availability of oxygen for the corrosion reactions; consequently, removing oxygen as one of the reactants should effectively eliminate corrosion. In terms of the waste tanks, excluding oxygen from the annular vault space, such as by continuous flushing with an inert gas, should substantially decrease corrosion of the external surfaces of the mild steel tanks (100% exclusion of oxygen is probably not practicable). Laboratory corrosion testing was conducted at Pacific Northwest National Laboratory to give a preliminary assessment of the ability of nitrogen-inerting to reduce steel corrosion. This report summarizes test results obtained after 18-month corrosion tests comparing open-quotes nitrogen-inertedclose quotes corrosion with open-quotes air-equilibratedclose quotes corrosion under simulated tank vault conditions

  3. PAIRWISE BLENDING OF HIGH LEVEL WASTE (HLW)

    Energy Technology Data Exchange (ETDEWEB)

    CERTA, P.J.

    2006-02-22

    The primary objective of this study is to demonstrate a mission scenario that uses pairwise and incidental blending of high level waste (HLW) to reduce the total mass of HLW glass. Secondary objectives include understanding how recent refinements to the tank waste inventory and solubility assumptions affect the mass of HLW glass and how logistical constraints may affect the efficacy of HLW blending.

  4. PAIRWISE BLENDING OF HIGH LEVEL WASTE

    International Nuclear Information System (INIS)

    CERTA, P.J.

    2006-01-01

    The primary objective of this study is to demonstrate a mission scenario that uses pairwise and incidental blending of high level waste (HLW) to reduce the total mass of HLW glass. Secondary objectives include understanding how recent refinements to the tank waste inventory and solubility assumptions affect the mass of HLW glass and how logistical constraints may affect the efficacy of HLW blending

  5. High-level-waste immobilization

    International Nuclear Information System (INIS)

    Crandall, J.L.

    1982-01-01

    Analysis of risks, environmental effects, process feasibility, and costs for disposal of immobilized high-level wastes in geologic repositories indicates that the disposal system safety has a low sensitivity to the choice of the waste disposal form

  6. Use of the Pipe ExplorerTM System to Deploy a Custom Gamma Tool in the Laterals Beneath High Level Waste Tanks in the 'A' and 'SX' Tank Farms, US DOE Hanford Site

    International Nuclear Information System (INIS)

    Kendrick, D.T.; Price, R.K.; Randall, R.R.; Myers, D.A.

    2006-01-01

    The 'laterals' are 3-inch tubing installed beneath single shell high level waste tanks in the 'A' and 'SX' Tank Farms at the US DOE Hanford Site during the late 1950's as part of a multifaceted leak detection system. A pneumatic deployment/wire line retrieval system was originally used to deploy two different custom Geiger-Muller detectors (a 'RED GM' and a 'GREEN GM') into the laterals for the purposes of characterizing activity levels in the soils beneath the waste tanks. Logging of the laterals was carried out from the mid 1970's through the early 1990's, when the activity was suspended. In support of the on-going vadose zone characterization efforts in the tank farms, CH2M Hill Hanford Group Inc. contracted with Apogen Technologies to utilize the Pipe Explorer TM system to deploy a custom gamma tool designed by Three Rivers Scientific and operated by Pacific Northwest Geophysics into selected laterals in the 'A' and 'SX' tank farms. The Pipe Explorer TM System is a unique deployment tool that utilizes a patented inverting membrane technology to deploy various detectors into piping, duct and drain lines. The conventional Pipe Explorer TM system was modified to interface with the PNG tool cabling and winch system that is typically used in bore hole applications. The gamma tool is comprised of three different detector systems, each with a different sensitivity. The most sensitive detector is a sodium iodide spectral gamma detector utilizing an on-board multi-channel analyzer. This detector is sensitive enough to measure the natural background radioactivity in these soils. Two additional Geiger-Muller gamma ray detectors complete the detector complement of the tool. These were designed with sensitivities similar to the historically used 'Green' and 'Red' GM detectors. The detectors were calibrated for Cs-137 concentration in the formation, and incorporated a correction for gamma ray attenuation due to the steel pipe of the lateral. The calibrations are traceable to

  7. USE OF AN EQUILIBRIUM MODEL TO FORECAST DISSOLUTION EFFECTIVENESS, SAFETY IMPACTS, AND DOWNSTREAM PROCESSABILITY FROM OXALIC ACID AIDED SLUDGE REMOVAL IN SAVANNAH RIVER SITE HIGH LEVEL WASTE TANKS 1-15

    Energy Technology Data Exchange (ETDEWEB)

    KETUSKY, EDWARD

    2005-10-31

    This thesis details a graduate research effort written to fulfill the Magister of Technologiae in Chemical Engineering requirements at the University of South Africa. The research evaluates the ability of equilibrium based software to forecast dissolution, evaluate safety impacts, and determine downstream processability changes associated with using oxalic acid solutions to dissolve sludge heels in Savannah River Site High Level Waste (HLW) Tanks 1-15. First, a dissolution model is constructed and validated. Coupled with a model, a material balance determines the fate of hypothetical worst-case sludge in the treatment and neutralization tanks during each chemical adjustment. Although sludge is dissolved, after neutralization more is created within HLW. An energy balance determines overpressurization and overheating to be unlikely. Corrosion induced hydrogen may overwhelm the purge ventilation. Limiting the heel volume treated/acid added and processing the solids through vitrification is preferred and should not significantly increase the number of glass canisters.

  8. High-Level Waste Melter Study Report

    Energy Technology Data Exchange (ETDEWEB)

    Perez Jr, Joseph M; Bickford, Dennis F; Day, Delbert E; Kim, Dong-Sang; Lambert, Steven L; Marra, Sharon L; Peeler, David K; Strachan, Denis M; Triplett, Mark B; Vienna, John D; Wittman, Richard S

    2001-07-13

    At the Hanford Site in Richland, Washington, the path to site cleanup involves vitrification of the majority of the wastes that currently reside in large underground tanks. A Joule-heated glass melter is the equipment of choice for vitrifying the high-level fraction of these wastes. Even though this technology has general national and international acceptance, opportunities may exist to improve or change the technology to reduce the enormous cost of accomplishing the mission of site cleanup. Consequently, the U.S. Department of Energy requested the staff of the Tanks Focus Area to review immobilization technologies, waste forms, and modifications to requirements for solidification of the high-level waste fraction at Hanford to determine what aspects could affect cost reductions with reasonable long-term risk. The results of this study are summarized in this report.

  9. Evolution of chemical conditions and estimated solubility controls on radionuclides in the residual waste layer during post-closure aging of high-level waste tanks

    Energy Technology Data Exchange (ETDEWEB)

    Denham, M. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Millings, M. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2012-08-28

    This document provides information specific to H-Area waste tanks that enables a flow and transport model with limited chemical capabilities to account for varying waste release from the tanks through time. The basis for varying waste release is solubilities of radionuclides that change as pore fluids passing through the waste change in composition. Pore fluid compositions in various stages were generated by simulations of tank grout degradation. The first part of the document describes simulations of the degradation of the reducing grout in post-closure tanks. These simulations assume flow is predominantly through a water saturated porous medium. The infiltrating fluid that reacts with the grout is assumed to be fluid that has passed through the closure cap and into the tank. The results are three stages of degradation referred to as Reduced Region II, Oxidized Region II, and Oxidized Region III. A reaction path model was used so that the transitions between each stage are noted by numbers of pore volumes of infiltrating fluid reacted. The number of pore volumes to each transition can then be converted to time within a flow and transport model. The bottoms of some tanks in H-Area are below the water table requiring a different conceptual model for grout degradation. For these simulations the reacting fluid was assumed to be 10% infiltrate through the closure cap and 90% groundwater. These simulations produce an additional four pore fluid compositions referred to as Conditions A through D and were intended to simulate varying degrees of groundwater influence. The most probable degradation path for the submerged tanks is Condition C to Condition D to Oxidized Region III and eventually to Condition A. Solubilities for Condition A are estimated in the text for use in sensitivity analyses if needed. However, the grout degradation simulations did not include sufficient pore volumes of infiltrating fluid for the grout to evolve to Condition A. Solubility controls for use

  10. High-Level Waste (HLW) Feed Process Control Strategy

    International Nuclear Information System (INIS)

    STAEHR, T.W.

    2000-01-01

    The primary purpose of this document is to describe the overall process control strategy for monitoring and controlling the functions associated with the Phase 1B high-level waste feed delivery. This document provides the basis for process monitoring and control functions and requirements needed throughput the double-shell tank system during Phase 1 high-level waste feed delivery. This document is intended to be used by (1) the developers of the future Process Control Plan and (2) the developers of the monitoring and control system

  11. Development of technical information database for high level waste disposal

    International Nuclear Information System (INIS)

    Kudo, Koji; Takada, Susumu; Kawanishi, Motoi

    2005-01-01

    A concept design of the high level waste disposal information database and the disposal technologies information database are explained. The high level waste disposal information database contains information on technologies, waste, management and rules, R and D, each step of disposal site selection, characteristics of sites, demonstration of disposal technology, design of disposal site, application for disposal permit, construction of disposal site, operation and closing. Construction of the disposal technologies information system and the geological disposal technologies information system is described. The screen image of the geological disposal technologies information system is shown. User is able to search the full text retrieval and attribute retrieval in the image. (S.Y. )

  12. Inspection and evaluation of Nuclear Fuel Services high-level waste storage system. Program plan

    International Nuclear Information System (INIS)

    1980-01-01

    The Nuclear Regulatory Commission has expressed the need for additional information concerning the condition of the high-level waste tanks at the Western New York State Nuclear Service center near West Valley, New York. This information is intended to be used in evaluating the safety of continued storage and in the development of alternatives for final disposition of the high-level waste

  13. Multipurpose optimization models for high level waste vitrification

    International Nuclear Information System (INIS)

    Hoza, M.

    1994-08-01

    Optimal Waste Loading (OWL) models have been developed as multipurpose tools for high-level waste studies for the Tank Waste Remediation Program at Hanford. Using nonlinear programming techniques, these models maximize the waste loading of the vitrified waste and optimize the glass formers composition such that the glass produced has the appropriate properties within the melter, and the resultant vitrified waste form meets the requirements for disposal. The OWL model can be used for a single waste stream or for blended streams. The models can determine optimal continuous blends or optimal discrete blends of a number of different wastes. The OWL models have been used to identify the most restrictive constraints, to evaluate prospective waste pretreatment methods, to formulate and evaluate blending strategies, and to determine the impacts of variability in the wastes. The OWL models will be used to aid in the design of frits and the maximize the waste in the glass for High-Level Waste (HLW) vitrification

  14. Techniques for the solidification of high-level wastes

    International Nuclear Information System (INIS)

    1977-01-01

    The problem of the long-term management of the high-level wastes from the reprocessing of irradiated nuclear fuel is receiving world-wide attention. While the majority of the waste solutions from the reprocessing of commercial fuels are currently being stored in stainless-steel tanks, increasing effort is being devoted to developing technology for the conversion of these wastes into solids. A number of full-scale solidification facilities are expected to come into operation in the next decade. The object of this report is to survey and compare all the work currently in progress on the techniques available for the solidification of high-level wastes. It will examine the high-level liquid wastes arising from the various processes currently under development or in operation, the advantages and disadvantages of each process for different types and quantities of waste solutions, the stages of development, the scale-up potential and flexibility of the processes

  15. DEFENSE HIGH LEVEL WASTE GLASS DEGRADATION

    International Nuclear Information System (INIS)

    Ebert, W.

    2001-01-01

    The purpose of this Analysis/Model Report (AMR) is to document the analyses that were done to develop models for radionuclide release from high-level waste (HLW) glass dissolution that can be integrated into performance assessment (PA) calculations conducted to support site recommendation and license application for the Yucca Mountain site. This report was developed in accordance with the ''Technical Work Plan for Waste Form Degradation Process Model Report for SR'' (CRWMS M andO 2000a). It specifically addresses the item, ''Defense High Level Waste Glass Degradation'', of the product technical work plan. The AP-3.15Q Attachment 1 screening criteria determines the importance for its intended use of the HLW glass model derived herein to be in the category ''Other Factors for the Postclosure Safety Case-Waste Form Performance'', and thus indicates that this factor does not contribute significantly to the postclosure safety strategy. Because the release of radionuclides from the glass will depend on the prior dissolution of the glass, the dissolution rate of the glass imposes an upper bound on the radionuclide release rate. The approach taken to provide a bound for the radionuclide release is to develop models that can be used to calculate the dissolution rate of waste glass when contacted by water in the disposal site. The release rate of a particular radionuclide can then be calculated by multiplying the glass dissolution rate by the mass fraction of that radionuclide in the glass and by the surface area of glass contacted by water. The scope includes consideration of the three modes by which water may contact waste glass in the disposal system: contact by humid air, dripping water, and immersion. The models for glass dissolution under these contact modes are all based on the rate expression for aqueous dissolution of borosilicate glasses. The mechanism and rate expression for aqueous dissolution are adequately understood; the analyses in this AMR were conducted to

  16. Current high-level waste solidification technology

    International Nuclear Information System (INIS)

    Bonner, W.F.; Ross, W.A.

    1976-01-01

    Technology has been developed in the U.S. and abroad for solidification of high-level waste from nuclear power production. Several processes have been demonstrated with actual radioactive waste and are now being prepared for use in the commercial nuclear industry. Conversion of the waste to a glass form is favored because of its high degree of nondispersibility and safety

  17. Atmospheric Pressure Effect of Retained Gas in High Level Waste

    International Nuclear Information System (INIS)

    Weber, A.H.

    1999-01-01

    Isolated high level waste tanks in H-Area have unexplained changes in waste-level which have been attributed to environmental effects including pressure, temperature, and relative humidity. Previous studies at SRS have considered waste-level changes from causes not including the presence of gas in the salt cake. This study was undertaken to determine the effect of atmospheric pressure on gas in the salt cake and resultant changes in the supernate level of Tank 41H, and to model that effect if possible. A simple theory has been developed to account for changes in the supernate level in a high level waste tank containing damp salt cake as the response of trapped gases to changes in the ambient pressure. The gas is modeled as an ideal gas retained as bubbles within the interstitial spaces in the salt cake and distributed uniformly throughout the tank. The model does not account for consistent long term increases or decreases in the tank level. Any such trend in the tank level is attributed to changes in the liquid content in the tank (from condensation, evaporation, etc.) and is removed from the data prior to the void estimation. Short term fluctuations in the tank level are explained as the response of the entrained gas volume to changes in the ambient pressure. The model uses the response of the tank level to pressure changes to estimate an average void fraction for the time period of interest. This estimate of the void is then used to predict the expected level response. The theory was applied to three separate time periods of the level data for tank 41H as follows: (1) May 3, 1993 through August 3, 1993, (2) January 23, 1994 through April 21, 1994, and (3) June 4, 1994 through August 24, 1994. A strong correlation was found between fluctuations in the tank level and variations in the ambient pressure. This correlation is a clear marker of the presence of entrained gases in the tank. From model calculations, an average void fraction of 11 percent was estimated to

  18. Studies of Mercury in High Level Waste Systems

    Energy Technology Data Exchange (ETDEWEB)

    WILMARTH, WILLIAMR.

    2004-03-31

    Historically, mercury was added to the nuclear weapons processing as a catalyst for the dissolution of aluminum metal. After neutralization the mercury was disposed to the High Level Waste tanks where its speciation led to mercuric oxides/hydroxides in the sludge and a small soluble mercuric ion concentration in the alkaline supernate. This report in its original revision described a three-pronged approach for studying possible sources of elevated mercury vapor in and around the Tank Farm evaporator systems. This approach examined the engineering differences between the evaporator designs, the chemistry of mercury and its concentrations, and the potential for organomercury species.During the course of this work, the presence of dimethylmercury in the High Level Waste system was established. Vapor and liquid samples were analyzed from the three evaporator systems (2H, 2F, and 3H) along with the transfer lines leading to the Effluent Treatment Facility and tanks within the ETF. The magnitude of the dimethylmercury concentrations led to ventilation modifications at the 3H Evaporator and design for modifications at the 2H evaporator. Continued sampling efforts are aimed at understanding the boundaries of where dimethylmercury vapors exist.

  19. High Level Waste System Impacts from Acid Dissolution of Sludge

    Energy Technology Data Exchange (ETDEWEB)

    KETUSKY, EDWARD

    2006-04-20

    This research evaluates the ability of OLI{copyright} equilibrium based software to forecast Savannah River Site High Level Waste system impacts from oxalic acid dissolution of Tank 1-15 sludge heels. Without further laboratory and field testing, only the use of oxalic acid can be considered plausible to support sludge heel dissolution on multiple tanks. Using OLI{copyright} and available test results, a dissolution model is constructed and validated. Material and energy balances, coupled with the model, identify potential safety concerns. Overpressurization and overheating are shown to be unlikely. Corrosion induced hydrogen could, however, overwhelm the tank ventilation. While pH adjustment can restore the minimal hydrogen generation, resultant precipitates will notably increase the sludge volume. OLI{copyright} is used to develop a flowsheet such that additional sludge vitrification canisters and other negative system impacts are minimized. Sensitivity analyses are used to assess the processability impacts from variations in the sludge/quantities of acids.

  20. Immobilized high-level waste interim storage alternatives generation and analysis and decision report

    International Nuclear Information System (INIS)

    CALMUS, R.B.

    1999-01-01

    This report presents a study of alternative system architectures to provide onsite interim storage for the immobilized high-level waste produced by the Tank Waste Remediation System (TWRS) privatization vendor. It examines the contract and program changes that have occurred and evaluates their impacts on the baseline immobilized high-level waste (IHLW) interim storage strategy. In addition, this report documents the recommended initial interim storage architecture and implementation path forward

  1. High-level waste management technology program plan

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, H.D.

    1995-01-01

    The purpose of this plan is to document the integrated technology program plan for the Savannah River Site (SRS) High-Level Waste (HLW) Management System. The mission of the SRS HLW System is to receive and store SRS high-level wastes in a see and environmentally sound, and to convert these wastes into forms suitable for final disposal. These final disposal forms are borosilicate glass to be sent to the Federal Repository, Saltstone grout to be disposed of on site, and treated waste water to be released to the environment via a permitted outfall. Thus, the technology development activities described herein are those activities required to enable successful accomplishment of this mission. The technology program is based on specific needs of the SRS HLW System and organized following the systems engineering level 3 functions. Technology needs for each level 3 function are listed as reference, enhancements, and alternatives. Finally, FY-95 funding, deliverables, and schedules are s in Chapter IV with details on the specific tasks that are funded in FY-95 provided in Appendix A. The information in this report represents the vision of activities as defined at the beginning of the fiscal year. Depending on emergent issues, funding changes, and other factors, programs and milestones may be adjusted during the fiscal year. The FY-95 SRS HLW technology program strongly emphasizes startup support for the Defense Waste Processing Facility and In-Tank Precipitation. Closure of technical issues associated with these operations has been given highest priority. Consequently, efforts on longer term enhancements and alternatives are receiving minimal funding. However, High-Level Waste Management is committed to participation in the national Radioactive Waste Tank Remediation Technology Focus Area. 4 refs., 5 figs., 9 tabs.

  2. High-level waste management technology program plan

    International Nuclear Information System (INIS)

    Harmon, H.D.

    1995-01-01

    The purpose of this plan is to document the integrated technology program plan for the Savannah River Site (SRS) High-Level Waste (HLW) Management System. The mission of the SRS HLW System is to receive and store SRS high-level wastes in a see and environmentally sound, and to convert these wastes into forms suitable for final disposal. These final disposal forms are borosilicate glass to be sent to the Federal Repository, Saltstone grout to be disposed of on site, and treated waste water to be released to the environment via a permitted outfall. Thus, the technology development activities described herein are those activities required to enable successful accomplishment of this mission. The technology program is based on specific needs of the SRS HLW System and organized following the systems engineering level 3 functions. Technology needs for each level 3 function are listed as reference, enhancements, and alternatives. Finally, FY-95 funding, deliverables, and schedules are s in Chapter IV with details on the specific tasks that are funded in FY-95 provided in Appendix A. The information in this report represents the vision of activities as defined at the beginning of the fiscal year. Depending on emergent issues, funding changes, and other factors, programs and milestones may be adjusted during the fiscal year. The FY-95 SRS HLW technology program strongly emphasizes startup support for the Defense Waste Processing Facility and In-Tank Precipitation. Closure of technical issues associated with these operations has been given highest priority. Consequently, efforts on longer term enhancements and alternatives are receiving minimal funding. However, High-Level Waste Management is committed to participation in the national Radioactive Waste Tank Remediation Technology Focus Area. 4 refs., 5 figs., 9 tabs

  3. Permitting plan for the high-level waste interim storage

    Energy Technology Data Exchange (ETDEWEB)

    Deffenbaugh, M.L.

    1997-04-23

    This document addresses the environmental permitting requirements for the transportation and interim storage of solidified high-level waste (HLW) produced during Phase 1 of the Hanford Site privatization effort. Solidified HLW consists of canisters containing vitrified HLW (glass) and containers that hold cesium separated during low-level waste pretreatment. The glass canisters and cesium containers will be transported to the Canister Storage Building (CSB) in a U.S. Department of Energy (DOE)-provided transportation cask via diesel-powered tractor trailer. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage of Tank Waste Remediation Systems (TWRS) immobilized HLW (IHLW) and other canistered high-level waste forms; and (2) interim storage and disposal of TWRS immobilized low-activity tank waste (ILAW). An environmental requirements checklist and narrative was developed to identify the permitting path forward for the HLW interim storage (HLWIS) project (See Appendix B). This permitting plan will follow the permitting logic developed in that checklist.

  4. Permitting plan for the high-level waste interim storage

    International Nuclear Information System (INIS)

    Deffenbaugh, M.L.

    1997-01-01

    This document addresses the environmental permitting requirements for the transportation and interim storage of solidified high-level waste (HLW) produced during Phase 1 of the Hanford Site privatization effort. Solidified HLW consists of canisters containing vitrified HLW (glass) and containers that hold cesium separated during low-level waste pretreatment. The glass canisters and cesium containers will be transported to the Canister Storage Building (CSB) in a U.S. Department of Energy (DOE)-provided transportation cask via diesel-powered tractor trailer. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage of Tank Waste Remediation Systems (TWRS) immobilized HLW (IHLW) and other canistered high-level waste forms; and (2) interim storage and disposal of TWRS immobilized low-activity tank waste (ILAW). An environmental requirements checklist and narrative was developed to identify the permitting path forward for the HLW interim storage (HLWIS) project (See Appendix B). This permitting plan will follow the permitting logic developed in that checklist

  5. Glass formulation for phase 1 high-level waste vitrification

    International Nuclear Information System (INIS)

    Vienna, J.D.; Hrma, P.R.

    1996-04-01

    The purpose of this study is to provide potential glass formulations for prospective Phase 1 High-Level Waste (HLW) vitrification at Hanford. The results reported here will be used to aid in developing a Phase 1 HLW vitrification request for proposal (RFP) and facilitate the evaluation of ensuing proposals. The following factors were considered in the glass formulation effort: impact on total glass volume of requiring the vendor to process each of the tank compositions independently versus as a blend; effects of imposing typical values of B 2 O 3 content and waste loading in HLW borosilicate glasses as restrictions on the vendors (according to WAPS 1995, the typical values are 5--10 wt% B 2 O 3 and 20--40 wt% waste oxide loading); impacts of restricting the processing temperature to 1,150 C on eventual glass volume; and effects of caustic washing on any of the selected tank wastes relative to glass volume

  6. Application of new technologies for characterization of Hanford Site high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Winters, W.I.

    1998-02-03

    To support remediation of Hanford Site high-level radioactive waste tanks, new chemical and physical measurement technologies must be developed and deployed. This is a major task of the Chemistry Analysis Technology Support (CATS) group of the Hanford Corporation. New measurement methods are required for efficient and economical resolution of tank waste safety, waste retrieval, and disposal issues. These development and deployment activities are performed in cooperation with Waste Management Federal Services of Hanford, Inc. This paper provides an overview of current analytical technologies in progress. The high-level waste at the Hanford Site is chemically complex because of the numerous processes used in past nuclear fuel reprocessing there, and a variety of technologies is required for effective characterization. Programmatic and laboratory operational needs drive the selection of new technologies for characterizing Hanford Site high-level waste, and these technologies are developed for deployment in laboratories, hot cells or in the field. New physical methods, such as the propagating reactive systems screening tool (PRSST) to measure the potential for self-propagating reactions in stored wastes, are being implemented. Technology for sampling and measuring gases trapped within the waste matrix is being used to evaluate flammability hazards associated with gas releases from stored wastes. Application of new inductively coupled plasma and laser ablation mass spectrometry systems at the Hanford Site`s 222-S Laboratory will be described. A Raman spectroscopy probe mounted in a cone penetrometer to measure oxyanions in wastes or soils will be described. The Hanford Site has used large volumes of organic complexants and acids in processing waste, and capillary zone electrophoresis (CZE) methods have been developed for determining several of the major organic components in complex waste tank matrices. The principles involved, system installation, and results from

  7. Ramifications of defining high-level waste

    International Nuclear Information System (INIS)

    Wood, D.E.; Campbell, M.H.; Shupe, M.W.

    1987-01-01

    The Nuclear Regulatory Commission (NRC) is considering rule making to provide a concentration-based definition of high-level waste (HLW) under authority derived from the Nuclear Waste Policy Act (NWPA) of 1982 and the Low Level Waste Policy Amendments Act of 1985. The Department of Energy (DOE), which has the responsibility to dispose of certain kinds of commercial waste, is supporting development of a risk-based classification system by the Oak Ridge National Laboratory to assist in developing and implementing the NRC rule. The system is two dimensional, with the axes based on the phrases highly radioactive and requires permanent isolation in the definition of HLW in the NWPA. Defining HLW will reduce the ambiguity in the present source-based definition by providing concentration limits to establish which materials are to be called HLW. The system allows the possibility of greater-confinement disposal for some wastes which do not require the degree of isolation provided by a repository. The definition of HLW will provide a firm basis for waste processing options which involve partitioning of waste into a high-activity stream for repository disposal, and a low-activity stream for disposal elsewhere. Several possible classification systems have been derived and the characteristics of each are discussed. The Defense High Level Waste Technology Lead Office at DOE - Richland Operations Office, supported by Rockwell Hanford Operations, has coordinated reviews of the ORNL work by a technical peer review group and other DOE offices. The reviews produced several recommendations and identified several issues to be addressed in the NRC rule making. 10 references, 3 figures

  8. Radiation transport in high-level waste form

    International Nuclear Information System (INIS)

    Arakali, V.S.; Barnes, S.M.

    1992-01-01

    The waste form selected for vitrifying high-level nuclear waste stored in underground tanks at West Valley, NY is borosilicate glass. The maximum radiation level at the surface of a canister filled with the high-level waste form is prescribed by repository design criteria for handling and disposition of the vitrified waste. This paper presents an evaluation of the radiation transport characteristics for the vitreous waste form expected to be produced at West Valley and the resulting neutron and gamma dose rates. The maximum gamma and neutron dose rates are estimated to be less than 7500 R/h and 10 mRem/h respectively at the surface of a West Valley canister filled with borosilicate waste glass

  9. High level waste fixation in cermet form

    International Nuclear Information System (INIS)

    Kobisk, E.H.; Aaron, W.S.; Quinby, T.C.; Ramey, D.W.

    1981-01-01

    Commercial and defense high level waste fixation in cermet form is being studied by personnel of the Isotopes Research Materials Laboratory, Solid State Division (ORNL). As a corollary to earlier research and development in forming high density ceramic and cermet rods, disks, and other shapes using separated isotopes, similar chemical and physical processing methods have been applied to synthetic and real waste fixation. Generally, experimental products resulting from this approach have shown physical and chemical characteristics which are deemed suitable for long-term storage, shipping, corrosive environments, high temperature environments, high waste loading, decay heat dissipation, and radiation damage. Although leach tests are not conclusive, what little comparative data are available show cermet to withstand hydrothermal conditions in water and brine solutions. The Soxhlet leach test, using radioactive cesium as a tracer, showed that leaching of cermet was about X100 less than that of 78 to 68 glass. Using essentially uncooled, untreated waste, cermet fixation was found to accommodate up to 75% waste loading and yet, because of its high thermal conductivity, a monolith of 0.6 m diameter and 3.3 m-length would have only a maximum centerline temperature of 29 K above the ambient value

  10. Decontamination of high-level waste canisters

    International Nuclear Information System (INIS)

    Nesbitt, J.F.; Slate, S.C.; Fetrow, L.K.

    1980-12-01

    This report presents evaluations of several methods for the in-process decontamination of metallic canisters containing any one of a number of solidified high-level waste (HLW) forms. The use of steam-water, steam, abrasive blasting, electropolishing, liquid honing, vibratory finishing and soaking have been tested or evaluated as potential techniques to decontaminate the outer surfaces of HLW canisters. Either these techniques have been tested or available literature has been examined to assess their applicability to the decontamination of HLW canisters. Electropolishing has been found to be the most thorough method to remove radionuclides and other foreign material that may be deposited on or in the outer surface of a canister during any of the HLW processes. Steam or steam-water spraying techniques may be adequate for some applications but fail to remove all contaminated forms that could be present in some of the HLW processes. Liquid honing and abrasive blasting remove contamination and foreign material very quickly and effectively from small areas and components although these blasting techniques tend to disperse the material removed from the cleaned surfaces. Vibratory finishing is very capable of removing the bulk of contamination and foreign matter from a variety of materials. However, special vibratory finishing equipment would have to be designed and adapted for a remote process. Soaking techniques take long periods of time and may not remove all of the smearable contamination. If soaking involves pickling baths that use corrosive agents, these agents may cause erosion of grain boundaries that results in rough surfaces

  11. High-level waste melter alternatives assessment report

    International Nuclear Information System (INIS)

    Calmus, R.B.

    1995-02-01

    This document describes the Tank Waste Remediation System (TWRS) High-Level Waste (HLW) Program's (hereafter referred to as HLW Program) Melter Candidate Assessment Activity performed in fiscal year (FY) 1994. The mission of the TWRS Program is to store, treat, and immobilize highly radioactive Hanford Site waste (current and future tank waste and encapsulated strontium and cesium isotopic sources) in an environmentally sound, safe, and cost-effective manner. The goal of the HLW Program is to immobilize the HLW fraction of pretreated tank waste into a vitrified product suitable for interim onsite storage and eventual offsite disposal at a geologic repository. Preparation of the encapsulated strontium and cesium isotopic sources for final disposal is also included in the HLW Program. As a result of trade studies performed in 1992 and 1993, processes planned for pretreatment of tank wastes were modified substantially because of increasing estimates of the quantity of high-level and transuranic tank waste remaining after pretreatment. This resulted in substantial increases in needed vitrification plant capacity compared to the capacity of original Hanford Waste Vitrification Plant (HWVP). The required capacity has not been finalized, but is expected to be four to eight times that of the HWVP design. The increased capacity requirements for the HLW vitrification plant's melter prompted the assessment of candidate high-capacity HLW melter technologies to determine the most viable candidates and the required development and testing (D and T) focus required to select the Hanford Site HLW vitrification plant melter system. An assessment process was developed in early 1994. This document describes the assessment team, roles of team members, the phased assessment process and results, resulting recommendations, and the implementation strategy

  12. High-level waste melter alternatives assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Calmus, R.B.

    1995-02-01

    This document describes the Tank Waste Remediation System (TWRS) High-Level Waste (HLW) Program`s (hereafter referred to as HLW Program) Melter Candidate Assessment Activity performed in fiscal year (FY) 1994. The mission of the TWRS Program is to store, treat, and immobilize highly radioactive Hanford Site waste (current and future tank waste and encapsulated strontium and cesium isotopic sources) in an environmentally sound, safe, and cost-effective manner. The goal of the HLW Program is to immobilize the HLW fraction of pretreated tank waste into a vitrified product suitable for interim onsite storage and eventual offsite disposal at a geologic repository. Preparation of the encapsulated strontium and cesium isotopic sources for final disposal is also included in the HLW Program. As a result of trade studies performed in 1992 and 1993, processes planned for pretreatment of tank wastes were modified substantially because of increasing estimates of the quantity of high-level and transuranic tank waste remaining after pretreatment. This resulted in substantial increases in needed vitrification plant capacity compared to the capacity of original Hanford Waste Vitrification Plant (HWVP). The required capacity has not been finalized, but is expected to be four to eight times that of the HWVP design. The increased capacity requirements for the HLW vitrification plant`s melter prompted the assessment of candidate high-capacity HLW melter technologies to determine the most viable candidates and the required development and testing (D and T) focus required to select the Hanford Site HLW vitrification plant melter system. An assessment process was developed in early 1994. This document describes the assessment team, roles of team members, the phased assessment process and results, resulting recommendations, and the implementation strategy.

  13. Evaluation and selection of candidate high-level waste forms

    International Nuclear Information System (INIS)

    1982-03-01

    Seven candidate waste forms being developed under the direction of the Department of Energy's National High-Level Waste (HLW) Technology Program, were evaluated as potential media for the immobilization and geologic disposal of high-level nuclear wastes. The evaluation combined preliminary waste form evaluations conducted at DOE defense waste-sites and independent laboratories, peer review assessments, a product performance evaluation, and a processability analysis. Based on the combined results of these four inputs, two of the seven forms, borosilicate glass and a titanate based ceramic, SYNROC, were selected as the reference and alternative forms for continued development and evaluation in the National HLW Program. Both the glass and ceramic forms are viable candidates for use at each of the DOE defense waste-sites; they are also potential candidates for immobilization of commercial reprocessing wastes. This report describes the waste form screening process, and discusses each of the four major inputs considered in the selection of the two forms

  14. Determination of total cyanide in Hanford Site high-level wastes

    Energy Technology Data Exchange (ETDEWEB)

    Winters, W.I. [Westinghouse Hanford Co., Richland, WA (United States); Pool, K.H. [Pacific Northwest Lab., Richland, WA (United States)

    1994-05-01

    Nickel ferrocyanide compounds (Na{sub 2-x}Cs{sub x}NiFe (CN){sub 6}) were produced in a scavenging process to remove {sup 137}Cs from Hanford Site single-shell tank waste supernates. Methods for determining total cyanide in Hanford Site high-level wastes are needed for the evaluation of potential exothermic reactions between cyanide and oxidizers such as nitrate and for safe storage, processing, and management of the wastes in compliance with regulatory requirements. Hanford Site laboratory experience in determining cyanide in high-level wastes is summarized. Modifications were made to standard cyanide methods to permit improved handling of high-level waste samples and to eliminate interferences found in Hanford Site waste matrices. Interferences and associated procedure modifications caused by high nitrates/nitrite concentrations, insoluble nickel ferrocyanides, and organic complexants are described.

  15. Determination of total cyanide in Hanford Site high-level wastes

    International Nuclear Information System (INIS)

    Winters, W.I.; Pool, K.H.

    1994-05-01

    Nickel ferrocyanide compounds (Na 2-x Cs x NiFe (CN) 6 ) were produced in a scavenging process to remove 137 Cs from Hanford Site single-shell tank waste supernates. Methods for determining total cyanide in Hanford Site high-level wastes are needed for the evaluation of potential exothermic reactions between cyanide and oxidizers such as nitrate and for safe storage, processing, and management of the wastes in compliance with regulatory requirements. Hanford Site laboratory experience in determining cyanide in high-level wastes is summarized. Modifications were made to standard cyanide methods to permit improved handling of high-level waste samples and to eliminate interferences found in Hanford Site waste matrices. Interferences and associated procedure modifications caused by high nitrates/nitrite concentrations, insoluble nickel ferrocyanides, and organic complexants are described

  16. Characteristics of solidified high-level waste products

    International Nuclear Information System (INIS)

    1979-01-01

    The object of the report is to contribute to the establishment of a data bank for future preparation of codes of practice and standards for the management of high-level wastes. The work currently in progress on measuring the properties of solidified high-level wastes is being studied

  17. Solidification of Savannah River Plant high level waste

    International Nuclear Information System (INIS)

    Maher, R.; Shafranek, L.F.; Kelley, J.A.; Zeyfang, R.W.

    1981-11-01

    Authorization for construction of the Defense Waste Processing Facility (DWPF) is expected in FY 83. The optimum time for stage 2 authorization is about three years later. Detailed design and construction will require approximately five years for stage 1, with stage 2 construction completed about two to three years later. Production of canisters of waste glass would begin in 1988, and the existing backlog of high level waste sludge stored at SRP would be worked off by about the year 2000. Stage 2 operation could begin in 1990. The technology and engineering are ready for construction and eventual operation of the DWPF for immobilizing high level radioactive waste at Savannah River Plant (SRP). Proceeding with this project will provide the public, and the leadership of this country, with a crucial demonstration that a major quantity of existing high level nuclear wastes can be safely and permanently immobilized. Early demonstration will both expedite and facilitate rational decision making on this aspect of the nuclear program. Delay in providing these facilities will result in significant DOE expenditures at SRP for new tanks just for continued temporary storage of wastes, and would probably result in dissipation of the intellectual and planning momentum that has built up in developing the project

  18. Hanford high-level waste melter system evaluation data packages

    International Nuclear Information System (INIS)

    Elliott, M.L.; Shafer, P.J.; Lamar, D.A.; Merrill, R.A.; Grunewald, W.; Roth, G.; Tobie, W.

    1996-03-01

    The Tank Waste Remediation System is selecting a reference melter system for the Hanford High-Level Waste vitrification plant. A melter evaluation was conducted in FY 1994 to narrow down the long list of potential melter technologies to a few for testing. A formal evaluation was performed by a Melter Selection Working Group (MSWG), which met in June and August 1994. At the June meeting, MSWG evaluated 15 technologies and selected six for more thorough evaluation at the Aug. meeting. All 6 were variations of joule-heated or induction-heated melters. Between the June and August meetings, Hanford site staff and consultants compiled data packages for each of the six melter technologies as well as variants of the baseline technologies. Information was solicited from melter candidate vendors to supplement existing information. This document contains the data packages compiled to provide background information to MSWG in support of the evaluation of the six technologies. (A separate evaluation was performed by Fluor Daniel, Inc. to identify balance of plant impacts if a given melter system was selected.)

  19. High-Level Waste Systems Plan. Revision 7 (U)

    International Nuclear Information System (INIS)

    Brooke, J.N.; Gregory, M.V.; Paul, P.; Taylor, G.; Wise, F.E.; Davis, N.R.; Wells, M.N.

    1996-10-01

    This revision of the High-Level Waste (HLW) System Plan aligns SRS HLW program planning with the DOE Savannah River (DOE-SR) Ten Year Plan (QC-96-0005, Draft 8/6), which was issued in July 1996. The objective of the Ten Year Plan is to complete cleanup at most nuclear sites within the next ten years. The two key principles of the Ten Year Plan are to accelerate the reduction of the most urgent risks to human health and the environment and to reduce mortgage costs. Accordingly, this System Plan describes the HLW program that will remove HLW from all 24 old-style tanks, and close 20 of those tanks, by 2006 with vitrification of all HLW by 2018. To achieve these goals, the DWPF canister production rate is projected to climb to 300 canisters per year starting in FY06, and remain at that rate through the end of the program in FY18, (Compare that to past System Plans, in which DWPF production peaked at 200 canisters per year, and the program did not complete until 2026.) An additional $247M (FY98 dollars) must be made available as requested over the ten year planning period, including a one-time $10M to enhance Late Wash attainment. If appropriate resources are made available, facility attainment issues are resolved and regulatory support is sufficient, then completion of the HLW program in 2018 would achieve a $3.3 billion cost savings to DOE, versus the cost of completing the program in 2026. Facility status information is current as of October 31, 1996

  20. National high-level waste systems analysis report

    Energy Technology Data Exchange (ETDEWEB)

    Kristofferson, K.; Oholleran, T.P.; Powell, R.H.

    1995-09-01

    This report documents the assessment of budgetary impacts, constraints, and repository availability on the storage and treatment of high-level waste and on both existing and pending negotiated milestones. The impacts of the availabilities of various treatment systems on schedule and throughput at four Department of Energy sites are compared to repository readiness in order to determine the prudent application of resources. The information modeled for each of these sites is integrated with a single national model. The report suggests a high-level-waste model that offers a national perspective on all high-level waste treatment and storage systems managed by the Department of Energy.

  1. High-level waste immobilization program: an overview

    International Nuclear Information System (INIS)

    Bonner, W.R.

    1979-09-01

    The High-Level Waste Immobilization Program is providing technology to allow safe, affordable immobilization and disposal of nuclear waste. Waste forms and processes are being developed on a schedule consistent with national needs for immobilization of high-level wastes stored at Savannah River, Hanford, Idaho National Engineering Laboratory, and West Valley, New York. This technology is directly applicable to high-level wastes from potential reprocessing of spent nuclear fuel. The program is removing one more obstacle previously seen as a potential restriction on the use and further development of nuclear power, and is thus meeting a critical technological need within the national objective of energy independence

  2. Handling and storage of conditioned high-level wastes

    International Nuclear Information System (INIS)

    1983-01-01

    This report deals with certain aspects of the management of one of the most important wastes, i.e. the handling and storage of conditioned (immobilized and packaged) high-level waste from the reprocessing of spent nuclear fuel and, although much of the material presented here is based on information concerning high-level waste from reprocessing LWR fuel, the principles, as well as many of the details involved, are applicable to all fuel types. The report provides illustrative background material on the arising and characteristics of high-level wastes and, qualitatively, their requirements for conditioning. The report introduces the principles important in conditioned high-level waste storage and describes the types of equipment and facilities, used or studied, for handling and storage of such waste. Finally, it discusses the safety and economic aspects that are considered in the design and operation of handling and storage facilities

  3. Handling and storage of conditioned high-level wastes

    International Nuclear Information System (INIS)

    Heafield, W.

    1984-01-01

    This paper deals with certain aspects of the management of one of the most important radioactive wastes arising from the nuclear fuel cycle, i.e. the handling and storage of conditioned high-level wastes. The paper is based on an IAEA report of the same title published during 1983 in the Technical Reports Series. The paper provides illustrative background material on the characteristics of high-level wastes and, qualitatively, their requirements for conditioning. The principles important in the storage of high-level wastes are reviewed in conjunction with the radiological and socio-political considerations involved. Four fundamentally different storage concepts are described with reference to published information and the safety aspects of particular storage concepts are discussed. Finally, overall conclusions are presented which confirm the availability of technology for constructing and operating conditioned high-level waste storage facilities for periods of at least several decades. (author)

  4. Development of melt compositions for sulphate bearing high level waste

    International Nuclear Information System (INIS)

    Jahagirdar, P.B.; Wattal, P.K.

    1997-09-01

    The report deals with the development and characterization of vitreous matrices for sulphate bearing high level waste. Studies were conducted in sodium borosilicate and lead borosilicate systems with the introduction of CaO, BaO, MgO etc. Lead borosilicate system was found to be compatible with sulphate bearing high level wastes. Detailed product evaluation carried on selected formulations is also described. (author)

  5. High-Level Waste System Process Interface Description

    International Nuclear Information System (INIS)

    D'Entremont, P.D.

    1999-01-01

    The High-Level Waste System is a set of six different processes interconnected by pipelines. These processes function as one large treatment plant that receives, stores, and treats high-level wastes from various generators at SRS and converts them into forms suitable for final disposal. The three major forms are borosilicate glass, which will be eventually disposed of in a Federal Repository, Saltstone to be buried on site, and treated water effluent that is released to the environment

  6. High-Level Waste System Process Interface Description

    Energy Technology Data Exchange (ETDEWEB)

    d' Entremont, P.D.

    1999-01-14

    The High-Level Waste System is a set of six different processes interconnected by pipelines. These processes function as one large treatment plant that receives, stores, and treats high-level wastes from various generators at SRS and converts them into forms suitable for final disposal. The three major forms are borosilicate glass, which will be eventually disposed of in a Federal Repository, Saltstone to be buried on site, and treated water effluent that is released to the environment.

  7. Alternatives generation and analysis for the phase 1 high-level waste pretreatment process selection

    International Nuclear Information System (INIS)

    Manuel, A.F.

    1997-01-01

    This report evaluates the effects of enhanced sludge washing and sludge washing without caustic leaching during the preparation of the Phase 1 high-level waste feeds. The pretreatment processing alternatives are evaluated against their ability to satisfy contractual, cost minimization, and other criteria. The information contained in this report is consistent with, and supplemental to, the Tank Waste Remediation System Operation and Utilization Plan (Kirkbride et al. 1997)

  8. Alternatives generation and analysis for the phase 1 high-level waste pretreatment process selection

    Energy Technology Data Exchange (ETDEWEB)

    Manuel, A.F.

    1997-10-02

    This report evaluates the effects of enhanced sludge washing and sludge washing without caustic leaching during the preparation of the Phase 1 high-level waste feeds. The pretreatment processing alternatives are evaluated against their ability to satisfy contractual, cost minimization, and other criteria. The information contained in this report is consistent with, and supplemental to, the Tank Waste Remediation System Operation and Utilization Plan (Kirkbride et al. 1997).

  9. Advanced High-Level Waste Glass Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Peeler, David K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vienna, John D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schweiger, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fox, Kevin M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-01

    The U.S. Department of Energy Office of River Protection (ORP) has implemented an integrated program to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. The integrated ORP program is focused on providing a technical, science-based foundation from which key decisions can be made regarding the successful operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) facilities. The fundamental data stemming from this program will support development of advanced glass formulations, key process control models, and tactical processing strategies to ensure safe and successful operations for both the low-activity waste (LAW) and high-level waste (HLW) vitrification facilities with an appreciation toward reducing overall mission life. The purpose of this advanced HLW glass research and development plan is to identify the near-, mid-, and longer-term research and development activities required to develop and validate advanced HLW glasses and their associated models to support facility operations at WTP, including both direct feed and full pretreatment flowsheets. This plan also integrates technical support of facility operations and waste qualification activities to show the interdependence of these activities with the advanced waste glass (AWG) program to support the full WTP mission. Figure ES-1 shows these key ORP programmatic activities and their interfaces with both WTP facility operations and qualification needs. The plan is a living document that will be updated to reflect key advancements and mission strategy changes. The research outlined here is motivated by the potential for substantial economic benefits (e.g., significant increases in waste throughput and reductions in glass volumes) that will be realized when advancements in glass formulation continue and models supporting facility operations are implemented. Developing and applying advanced

  10. Managing the high level waste nuclear regulatory commission licensing process

    International Nuclear Information System (INIS)

    Baskin, K.P.

    1992-01-01

    This paper reports that the process for obtaining Nuclear Regulatory Commission permits for the high level waste storage facility is basically the same process commercial nuclear power plants followed to obtain construction permits and operating licenses for their facilities. Therefore, the experience from licensing commercial reactors can be applied to the high level waste facility. Proper management of the licensing process will be the key to the successful project. The management of the licensing process was categorized into four areas as follows: responsibility, organization, communication and documentation. Drawing on experience from nuclear power plant licensing and basic management principles, the management requirement for successfully accomplishing the project goals are discussed

  11. Treatment technologies for non-high-level wastes (USA)

    International Nuclear Information System (INIS)

    Cooley, C.R.; Clark, D.E.

    1976-06-01

    Non-high-level waste arising from operations at nuclear reactors, fuel fabrication facilities, and reprocessing facilities can be treated using one of several technical alternatives prior to storage. Each alternative and the associated experience and status of development are summarized. The technology for treating non-high-level wastes is generally available for industrial use. Improved techniques applicable to the commercial nuclear fuel cycle are being developed and demonstrated to reduce the volume of waste and to immobilize it for storage. 36 figures, 59 references

  12. An approximate-reasoning-based method for screening flammable gas tanks

    Energy Technology Data Exchange (ETDEWEB)

    Eisenhawer, S.W.; Bott, T.F.; Smith, R.E.

    1998-03-01

    High-level waste (HLW) produces flammable gases as a result of radiolysis and thermal decomposition of organics. Under certain conditions, these gases can accumulate within the waste for extended periods and then be released quickly into the dome space of the storage tank. As part of the effort to reduce the safety concerns associated with flammable gas in HLW tanks at Hanford, a flammable gas watch list (FGWL) has been established. Inclusion on the FGWL is based on criteria intended to measure the risk associated with the presence of flammable gas. It is important that all high-risk tanks be identified with high confidence so that they may be controlled. Conversely, to minimize operational complexity, the number of tanks on the watchlist should be reduced as near to the true number of flammable risk tanks as the current state of knowledge will support. This report presents an alternative to existing approaches for FGWL screening based on the theory of approximate reasoning (AR) (Zadeh 1976). The AR-based model emulates the inference process used by an expert when asked to make an evaluation. The FGWL model described here was exercised by performing two evaluations. (1) A complete tank evaluation where the entire algorithm is used. This was done for two tanks, U-106 and AW-104. U-106 is a single shell tank with large sludge and saltcake layers. AW-104 is a double shell tank with over one million gallons of supernate. Both of these tanks had failed the screening performed by Hodgson et al. (2) Partial evaluations using a submodule for the predictor likelihood for all of the tanks on the FGWL that had been flagged previously by Whitney (1995).

  13. An approximate-reasoning-based method for screening flammable gas tanks

    International Nuclear Information System (INIS)

    Eisenhawer, S.W.; Bott, T.F.; Smith, R.E.

    1998-03-01

    High-level waste (HLW) produces flammable gases as a result of radiolysis and thermal decomposition of organics. Under certain conditions, these gases can accumulate within the waste for extended periods and then be released quickly into the dome space of the storage tank. As part of the effort to reduce the safety concerns associated with flammable gas in HLW tanks at Hanford, a flammable gas watch list (FGWL) has been established. Inclusion on the FGWL is based on criteria intended to measure the risk associated with the presence of flammable gas. It is important that all high-risk tanks be identified with high confidence so that they may be controlled. Conversely, to minimize operational complexity, the number of tanks on the watchlist should be reduced as near to the true number of flammable risk tanks as the current state of knowledge will support. This report presents an alternative to existing approaches for FGWL screening based on the theory of approximate reasoning (AR) (Zadeh 1976). The AR-based model emulates the inference process used by an expert when asked to make an evaluation. The FGWL model described here was exercised by performing two evaluations. (1) A complete tank evaluation where the entire algorithm is used. This was done for two tanks, U-106 and AW-104. U-106 is a single shell tank with large sludge and saltcake layers. AW-104 is a double shell tank with over one million gallons of supernate. Both of these tanks had failed the screening performed by Hodgson et al. (2) Partial evaluations using a submodule for the predictor likelihood for all of the tanks on the FGWL that had been flagged previously by Whitney (1995)

  14. Screening the Hanford tanks for trapped gas

    Energy Technology Data Exchange (ETDEWEB)

    Whitney, P.

    1995-10-01

    The Hanford Site is home to 177 large, underground nuclear waste storage tanks. Hydrogen gas is generated within the waste in these tanks. This document presents the results of a screening of Hanford`s nuclear waste storage tanks for the presence of gas trapped in the waste. The method used for the screening is to look for an inverse correlation between waste level measurements and ambient atmospheric pressure. If the waste level in a tank decreases with an increase in ambient atmospheric pressure, then the compressibility may be attributed to gas trapped within the waste. In this report, this methodology is not used to estimate the volume of gas trapped in the waste. The waste level measurements used in this study were made primarily to monitor the tanks for leaks and intrusions. Four measurement devices are widely used in these tanks. Three of these measure the level of the waste surface. The remaining device measures from within a well embedded in the waste, thereby monitoring the liquid level even if the liquid level is below a dry waste crust. In the past, a steady rise in waste level has been taken as an indicator of trapped gas. This indicator is not part of the screening calculation described in this report; however, a possible explanation for the rise is given by the mathematical relation between atmospheric pressure and waste level used to support the screening calculation. The screening was applied to data from each measurement device in each tank. If any of these data for a single tank indicated trapped gas, that tank was flagged by this screening process. A total of 58 of the 177 Hanford tanks were flagged as containing trapped gas, including 21 of the 25 tanks currently on the flammable gas watch list.

  15. Evaluation of tank data for safety screening

    Energy Technology Data Exchange (ETDEWEB)

    SIMPSON, B.C.

    1999-07-22

    This document presents the evaluation of the adequacy of the sampling and analysis of Hanford tank wastes for safety screening. A comparison was made of the data collected through sampling and analysis to the data required by the Safety Screening Data Quality Objective (DQO) (Dukelow et al. 1995). The evaluation was for the purpose of determining whether the sampling met the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 93-5 Implementation Plan milestone 5.6.3.13. Milestone 5.6.3.13 states, ''Core sample all tanks by 2002'' (DOE-RL 1996). The milestone is considered completed for a specific tank if sufficient sample material was taken and analyzed to meet the Safety Screening DQO. A description of the scope is presented in Section 2.0. The definition of Safety Screening DQO needs is presented in Section 3.0. The logic used to determine if the sampling and analysis was sufficient is presented in Section 4.0. In Section 5.0, the tanks that satisfy the Safety Screening DQO are presented. In total, 138 tanks were identified as having been sampled since 1989, and 132 of those tanks met the established criteria. Six tanks did not meet the established criteria. Section 6.0 lists the 45 tanks that either did not meet the criteria or were not sampled since 1989.

  16. High level waste canister emplacement and retrieval concepts study

    International Nuclear Information System (INIS)

    1975-09-01

    Several concepts are described for the interim (20 to 30 years) storage of canisters containing high level waste, cladding waste, and intermediate level-TRU wastes. It includes requirements, ground rules and assumptions for the entire storage pilot plant. Concepts are generally evaluated and the most promising are selected for additional work. Follow-on recommendations are made

  17. High-Level waste process and product data annotated bibliography

    International Nuclear Information System (INIS)

    Stegen, G.E.

    1996-01-01

    The objective of this document is to provide information on available issued documents that will assist interested parties in finding available data on high-level waste and transuranic waste feed compositions, properties, behavior in candidate processing operations, and behavior on candidate product glasses made from those wastes. This initial compilation is only a partial list of available references

  18. Demonstrating Reliable High Level Waste Slurry Sampling Techniques to Support Hanford Waste Processing

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Steven E.

    2013-11-11

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capability using simulated Hanford High-Level Waste (HL W) formulations. This work represents one of the remaining technical issues with the high-level waste treatment mission at Hanford. The TOC must demonstrate the ability to adequately mix and sample high-level waste feed to meet the WTP Waste Acceptance Criteria and Data Quality Objectives. The sampling method employed must support both TOC and WTP requirements. To facilitate information transfer between the two facilities the mixing and sampling demonstrations are led by the One System Integrated Project Team. The One System team, Waste Feed Delivery Mixing and Sampling Program, has developed a full scale sampling loop to demonstrate sampler capability. This paper discusses the full scale sampling loops ability to meet precision and accuracy requirements, including lessons learned during testing. Results of the testing showed that the Isolok(R) sampler chosen for implementation provides precise, repeatable results. The Isolok(R) sampler accuracy as tested did not meet test success criteria. Review of test data and the test platform following testing by a sampling expert identified several issues regarding the sampler used to provide reference material used to judge the Isolok's accuracy. Recommendations were made to obtain new data to evaluate the sampler's accuracy utilizing a reference sampler that follows good sampling protocol.

  19. Long-term high-level waste technology. Composite quarterly technical report, January-March 1981

    International Nuclear Information System (INIS)

    Cornman, W.R.

    1981-08-01

    This composite quarterly technical report summarizes work performed at participating sites to immobilize high-level radioactive wastes. The report is structured along the lines of the Work Breakdown Structure adopted for use in the High-Level Waste Management Technology program. These are: (1) program management and support with subtasks of management and budget, environmental and safety assessments, and other support; (2) waste preparation with subtasks of in-situ storage or disposal, waste retrieval, and separation and concentration; (3) waste fixation with subtasks of waste form development and characterization, and process and equipment development; and (4) final handling with subtasks of canister development and characterization and onsite storage or disposal. Some of the highlights are: preliminary event trees defining possible accidents were completed in the safety assessment of continued in-tank storage of high-level waste at Hanford; two low-cost waste forms (tailored concrete and bitumen) were investigated as candidate immobilization forms at the Hanford in-situ disposal studies of high-level waste; in comparative impact tests at the same impact energy per specimen volume, the same mass of respirable sizes was observed at ANL for SRL Frit 131 glass, SYNROC B ceramic, and SYNROC D ceramic; leaching tests were conducted on alkoxide glasses; glass-ceramic, concrete, and SYNROC D; a process design description was written for the tailored ceramic process

  20. Experimental and analytical study for demonstration program on shielding of casks for high-level wastes

    International Nuclear Information System (INIS)

    Ueki, K.; Nakazawa, M.; Hattorl, S.; Ozaki, S.; Tamaki, H.; Kadotani, H.; Ishizuka, T.; Ishikawa, S.

    1993-01-01

    The following remarks were obtained from the experiment and the DOT 3.5 and the MCNP analyses on the gamma ray and the neutron dose equivalent rates in the cask of interest. 1. The cask has thinner neutron shielding parts around the trunnions. Significant neutrons streaming around the trunnion parts was observed which was also cleared by the MCNP analysis for the 252 Cf source experiment. Accordingly, detailed neutron streaming calculations are required to evaluate the dose levels around the trunnions when loading the vitrified high-level wastes. 2. The room-scattered obstructive neutrons, mainly originating from the neutrons penetrating around the trunnions, at the top and the bottom of the cask are reduced significantly by preparing the water tank at the top and the water layer at the bottom. Therefore, a more accurate experiment is to be carried out in the future shielding experiment especially for neutrons. However, because the water tank and the layer do not exist in the actual high-level wastes transport cask, the experiment without the water tank and layer are not dispensable to demonstrate the transport conditions of the actual cask, too. 3. The gamma-ray and the neutron dose equivalent rate distributions obtained from the DOT 3.5 and the MCNP calculations, respectively, agreed closely with the measured values in the cask areas of interest. Accordingly, the DOT 3.5 code and the MCNP code with the NESX estimator can be employed not only for the shielding analysis of the future experiments, but also for making a safety analysis report of high-level wastes transport casks. (J.P.N.)

  1. Crystal accumulation in the Hanford Waste Treatment Plant high level waste melter: Summary of 2017 experiments

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Fowley, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2018-01-11

    A full-scale, transparent mock-up of the Hanford Tank Waste Treatment and Immobilization Project High Level Waste glass melter riser and pour spout has been constructed to allow for testing with visual feedback of particle settling, accumulation, and resuspension when operating with a controlled fraction of crystals in the glass melt. Room temperature operation with silicone oil and magnetite particles simulating molten glass and spinel crystals, respectively, allows for direct observation of flow patterns and settling patterns. The fluid and particle mixture is recycled within the system for each test.

  2. Defense Waste Processing Facility: a remote process for solidification of Savannah River Plant high level waste

    International Nuclear Information System (INIS)

    Maher, R.; Shafranek, L.F.; Kelley, J.A.; Zeyfang, R.W.; Lethco, A.J.

    1982-03-01

    The Department of Energy is proposing that a Defense Waste Processing Facility be built at the Savannah River Plant (SRP) to remotely process and immobilize high level radioactive waste produced at the site. Research, development, and design of the facility is being provided by a multidisciplined task force of personnel from the Du Pont Company which designed, built and has operated SRP since 1950. This remotely operated facility will immobilize 28 million gallons of high level waste now stored in tanks, plus the waste to be generated from continued reprocessing operations. Borosilicate glass has been selected as the reference waste form for the immobilization process

  3. Transport of radioactive high level waste to the sun

    International Nuclear Information System (INIS)

    1972-01-01

    First results on the feasibility of the disposal of high-level waste from reprocessing plants for LWR fuer elements are presented. The emphasis is on the transport of the waste to the sun by meams of existing and planned carrier systems and on a mission-specific upper stage (AMOS). The article discusses transition orbits earth-sun, gravitation losses, suitable carrier systems, design criteria, types of propulsion (radioisojet with hydrogen as propulsion gas), safety aspects and costs. It arrives at the conclusion that the transport of radioactive high-level waste to the sun by means of spacecraft technology appears feasible for the future and that the costs are not too high. It is pointed out that this project can only be carried out in international cooperation. (HR) [de

  4. Development of glasses for high-level waste solidification

    International Nuclear Information System (INIS)

    Ross, W.A.; Mendel, J.E.

    1976-01-01

    In reviewing the characteristics of high-level waste and the desirable characteristics of a waste form it is apparent that glass castings offer a very good choice for waste disposal. Our current data shows that glass will provide the characteristics needed for disposal of high-level nuclear wastes. However, additional characterization work is being performed to extend our current data to actual waste and verify waste form behavior on the long term with accelerated tests

  5. The tracking of high level waste shipments-TRANSCOM system

    International Nuclear Information System (INIS)

    Johnson, P.E.; Joy, D.S.; Pope, R.B.

    1995-01-01

    The TRANSCOM (transportation tracking and communication) system is the U.S. Department of Energy's (DOE's) real-time system for tracking shipments of spent fuel, high-level wastes, and other high-visibility shipments of radioactive material. The TRANSCOM system has been operational since 1988. The system was used during FY1993 to track almost 100 shipments within the US.DOE complex, and it is accessed weekly by 10 to 20 users

  6. Evaluation of conditioned high-level waste forms

    International Nuclear Information System (INIS)

    Mendel, J.E.; Turcotte, R.P.; Chikalla, T.D.; Hench, L.L.

    1983-01-01

    The evaluation of conditioned high-level waste forms requires an understanding of radiation and thermal effects, mechanical properties, volatility, and chemical durability. As a result of nuclear waste research and development programs in many countries, a good understanding of these factors is available for borosilicate glass containing high-level waste. The IAEA through its coordinated research program has contributed to this understanding. Methods used in the evaluation of conditioned high-level waste forms are reviewed. In the US, this evaluation has been facilitated by the definition of standard test methods by the Materials Characterization Center (MCC), which was established by the Department of Energy (DOE) in 1979. The DOE has also established a 20-member Materials Review Board to peer-review the activities of the MCC. In addition to comparing waste forms, testing must be done to evaluate the behavior of waste forms in geologic repositories. Such testing is complex; accelerated tests are required to predict expected behavior for thousands of years. The tests must be multicomponent tests to ensure that all potential interactions between waste form, canister/overpack and corrosion products, backfill, intruding ground water and the repository rock, are accounted for. An overview of the status of such multicomponent testing is presented

  7. Overview of high-level waste management accomplishments

    International Nuclear Information System (INIS)

    Lawroski, H.; Berreth, J.R.; Freeby, W.A.

    1980-01-01

    Storage of power reactor spent fuel is necessary at present because of the lack of reprocessing operations particularly in the U.S. By considering the above solidification and storage scenario, there is more than reasonable assurance that acceptable, stable, low heat generation rate, solidified waste can be produced, and safely disposed. The public perception of no waste disposal solutions is being exploited by detractors of nuclear power application. The inability to even point to one overall system demonstration lends credibility to the negative assertions. By delaying the gathering of on-line information to qualify repository sites, and to implement a demonstration, the actions of the nuclear power detractors are self serving in that they can continue to point out there is no demonstration of satisfactory high-level waste disposal. By maintaining the liquid and solidified high-level waste in secure above ground storage until acceptable decay heat generation rates are achieved, by producing a compatible, high integrity, solid waste form, by providing a second or even third barrier as a compound container and by inserting the enclosed waste form in a qualified repository with spacing to assure moderately low temperature disposal conditions, there appears to be no technical reason for not progressing further with the disposal of high-level wastes and needed implementation of the complete nuclear power fuel cycle

  8. Defense High Level Waste Disposal Container System Description Document

    International Nuclear Information System (INIS)

    Pettit, N. E.

    2001-01-01

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms [IPWF]) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. US Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as co-disposal. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister inserted in the center and/or one or more DOE SNF canisters displacing a HLW canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by

  9. Long-term high-level waste technology. Composite report

    Science.gov (United States)

    Cornman, W. R.

    1981-12-01

    Research and development studies on the immobilization of high-level wastes from the chemical reprocessing of nuclear reactor fuels are summarized. The reports are grouped under the following tasks: (1) program management and support; (2) waste preparation; (3) waste fixation; and (4) final handling. Some of the highlights are: leaching properties were obtained for titanate and tailored ceramic materials being developed at ICPP to immobilize zirconia calcine; comparative leach tests, hot-cell tests, and process evaluations were conducted of waste form alternatives to borosilicate glass for the immobilization of SRP high-level wastes, experiments were run at ANL to qualify neutron activation analysis and radioactive tracers for measuring leach rates from simulated waste glasses; comparative leach test samples of SYNROC D were prepared, characterized, and tested at LLNL; encapsulation of glass marbles with lead or lead alloys was demonstrated on an engineering scale at PNL; a canister for reference Commercial HLW was designed at PNL; a study of the optimization of salt-crete was completed at SRL; a risk assessment showed that an investment for tornado dampers in the interim storage building of the DWPF is unjustified.

  10. Technical baseline description of high-level waste and low-activity waste feed mobilization and delivery

    International Nuclear Information System (INIS)

    Papp, I.G.

    1997-01-01

    This document is a compilation of information related to the high-level waste (HLW) and low-activity waste (LAW) feed staging, mobilization, and transfer/delivery issues. Information relevant to current Tank Waste Remediation System (TWRS) inventories and activities designed to feed the Phase I Privatization effort at the Hanford Site is included. Discussions on the higher level Phase II activities are offered for a perspective on the interfaces

  11. STUDIES OF POTENTIAL INHIBITORS OF SODIUM ALUMINOSILICATE SCALES IN HIGH-LEVEL WASTE EVAPORATION

    Energy Technology Data Exchange (ETDEWEB)

    Wilmarth, B; Lawrence Oji, L; Terri Fellinger, T; David Hobbs, D; Nilesh Badheka, N

    2008-02-27

    The Savannah River Site (SRS) has 49 underground storage tanks used to store High Level Waste (HLW). The tank space in these tanks must be managed to support the continued operation of key facilities. The reduction of the tank volumes in these tanks are accomplished through the use of three atmospheric pressure HLW evaporators. For a decade, evaporation of highly alkaline HLW containing aluminum and silicates has produced sodium aluminosilicate scales causing both operation and criticality hazards in the 2H Evaporator System. Segregation of aluminum-rich wastes from silicate-rich wastes minimizes the amount of scale produced and reduces cleaning expenses, but does not eliminate the scaling nor increases operation flexibility in waste process. Similar issues have affected the aluminum refining industry for many decades. Over the past several years, successful commercial products have been identified to eliminate aluminosilicate fouling in the aluminum industry, but have not been utilized in a nuclear environment. Laboratory quantities of three proprietary aluminosilicate scale inhibitors have been produced and been shown to prevent formation of scales. SRNL has been actively testing these potential inhibitors to examine their radiation stability, radiolytic degradation behaviors, and downstream impacts to determine their viability within the HLW system. One of the tested polymers successfully meets the established criteria for application in the nuclear environment. This paper will describe a summary of the methodology used to prioritize laboratory testing protocols based on potential impacts/risks identified for inhibitor deployment at SRS.

  12. Crystallization in high-level waste glass: A review of glass theory and noteworthy literature

    Energy Technology Data Exchange (ETDEWEB)

    Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-18

    There is a fundamental need to continue research aimed at understanding nepheline and spinel crystal formation in high-level waste (HLW) glass. Specifically, the formation of nepheline solids (K/NaAlSiO4) during slow cooling of HLW glass can reduce the chemical durability of the glass, which can cause a decrease in the overall durability of the glass waste form. The accumulation of spinel solids ((Fe, Ni, Mn, Zn)(Fe, Cr)2O4), while not detrimental to glass durability, can cause an array of processing problems inside HLW glass melters. In this review, the fundamental differences between glass and solid-crystals are explained using kinetic, thermodynamic, and viscosity arguments, and several highlights of glass-crystallization research, as it pertains to high-level waste vitrification, are described. In terms of mitigating spinel in the melter and both spinel and nepheline formation in the canister, the complexity of HLW glass and the intricate interplay between thermal, chemical, and kinetic factors further complicates this understanding. However, new experiments seeking to elucidate the contributing factors of crystal nucleation and growth in waste glass, and the compilation of data from older experiments, may go a long way towards helping to achieve higher waste loadings while developing more efficient processing strategies. Higher waste loadings and more efficient processing strategies will reduce the overall HLW Hanford Tank Waste Treatment and Immobilization Plant (WTP) vitrification facilities mission life.

  13. National high-level waste systems analysis plan

    International Nuclear Information System (INIS)

    Kristofferson, K.; Oholleran, T.P.; Powell, R.H.; Thiel, E.C.

    1995-05-01

    This document details the development of modeling capabilities that can provide a system-wide view of all US Department of Energy (DOE) high-level waste (HLW) treatment and storage systems. This model can assess the impact of budget constraints on storage and treatment system schedules and throughput. These impacts can then be assessed against existing and pending milestones to determine the impact to the overall HLW system. A nation-wide view of waste treatment availability will help project the time required to prepare HLW for disposal. The impacts of the availability of various treatment systems and throughput can be compared to repository readiness to determine the prudent application of resources or the need to renegotiate milestones

  14. Coupled processes in NRC high-level waste research

    International Nuclear Information System (INIS)

    Costanzi, F.A.

    1987-01-01

    The author discusses NRC research effort in support of evaluating license applications for disposal of nuclear waste and for promulgating regulations and issuing guidance documents on nuclear waste management. In order to do this they fund research activities at a number of laboratories, academic institutions, and commercial organizations. One of our research efforts is the coupled processes study. This paper discusses interest in coupled processes and describes the target areas of research efforts over the next few years. The specific research activities relate to the performance objectives of NRC's high-level waste (HLW) regulation and the U.S. Environmental Protection Agency (EPA) HLW standard. The general objective of the research program is to ensure the NRC has a sufficient independent technical base to make sound regulatory decisions

  15. High level waste (HLW) steam reducing station evaluation

    International Nuclear Information System (INIS)

    Gannon, R.E.

    1993-01-01

    Existing pressure equipment in High Level Waste does not have a documented technical baseline. Based on preliminary reviews, the existing equipment seems to be based on system required capacity instead of system capability. A planned approach to establish a technical baseline began September 1992 and used the Works Management System preventive maintenance schedule. Several issues with relief valves being undersized on steam reducing stations created a need to determine the risk of maintaining the steam in service. An Action Plan was developed to evaluate relief valves that did not have technical baselines and provided a path forward for continued operation. Based on Action Plan WER-HLE-931042, the steam systems will remain in service while the designs are being developed and implemented

  16. Powder technological vitrification of simulated high-level waste

    International Nuclear Information System (INIS)

    Gahlert, S.

    1988-03-01

    High-level waste simulate from the reprocessing of light water reactor and fast breeder fuel was vitrified by powder technology. After denitration with formaldehyde, the simulated HLW is mixed with glass frit and simultaneously dried in an oil-heated mixer. After 'in-can calcination' for at least 24 hours at 850 or 950 K (depending on the type of waste and glass), the mixture is hot-pressed in-can for several hours at 920 or 1020 K respectively, at pressures between 0.4 and 1.0 MPa. The technology has been demonstrated inactively up to diameters of 30 cm. Leach resistance is significantly enhanced when compared to common borosilicate glasses by the utilization of glasses with higher silicon and aluminium content and lower sodium content. (orig.) [de

  17. C-106 High-Level Waste Solids: Washing/Leaching and Solubility Versus Temperature Studies

    Energy Technology Data Exchange (ETDEWEB)

    GJ Lumetta; DJ Bates; PK Berry; JP Bramson; LP Darnell; OT Farmer III; LR Greenwood; FV Hoopes; RC Lettau; GF Piepel; CZ Soderquist; MJ Steele; RT Steele; MW Urie; JJ Wagner

    2000-01-26

    This report describes the results of a test conducted by Battelle to assess the effects of inhibited water washing and caustic leaching on the composition of the Hanford tank C-106 high-level waste (HLW) solids. The objective of this work was to determine the composition of the C-106 solids remaining after washing with 0.01M NaOH or leaching with 3M NaOH. Another objective of this test was to determine the solubility of various C-106 components as a function of temperature. The work was conducted according to test plan BNFL-TP-29953-8,Rev. 0, Determination of the Solubility of HLW Sludge Solids. The test went according to plan, with only minor deviations from the test plan. The deviations from the test plan are discussed in the experimental section.

  18. Glass Property Data and Models for Estimating High-Level Waste Glass Volume

    International Nuclear Information System (INIS)

    Vienna, John D.; Fluegel, Alexander; Kim, Dong-Sang; Hrma, Pavel R.

    2009-01-01

    This report describes recent efforts to develop glass property models that can be used to help estimate the volume of high-level waste (HLW) glass that will result from vitrification of Hanford tank waste. The compositions of acceptable and processable HLW glasses need to be optimized to minimize the waste-form volume and, hence, to save cost. A database of properties and associated compositions for simulated waste glasses was collected for developing property-composition models. This database, although not comprehensive, represents a large fraction of data on waste-glass compositions and properties that were available at the time of this report. Glass property-composition models were fit to subsets of the database for several key glass properties. These models apply to a significantly broader composition space than those previously publised. These models should be considered for interim use in calculating properties of Hanford waste glasses.

  19. Performance of a buried radioactive high level waste (HLW) glass after 24 years

    International Nuclear Information System (INIS)

    Jantzen, Carol M.; Kaplan, Daniel I.; Bibler, Ned E.; Peeler, David K.; John Plodinec, M.

    2008-01-01

    A radioactive high level waste glass was made in 1980 with Savannah River Site (SRS) Tank 15 waste. This glass was buried in a lysimeter in the SRS burial ground for 24 years. Lysimeter leachate data was available for the first 8 years. The glass was exhumed in 2004. The glass was predicted to be very durable and laboratory tests confirmed this. Scanning electron microscopy of the glass burial surface showed no significant glass alteration consistent with results of other laboratory and field tests. Radionuclide profiling for alpha, beta, and 137 Cs indicated that Pu was not enriched in the soil while 137 Cs and 9 deg. C Sr were enriched in the first few centimeters surrounding the glass. Lysimeter leachate data indicated that 9 deg. C Sr and 137 Cs leaching from the glass was diffusion controlled

  20. Glass Property Data and Models for Estimating High-Level Waste Glass Volume

    Energy Technology Data Exchange (ETDEWEB)

    Vienna, John D.; Fluegel, Alexander; Kim, Dong-Sang; Hrma, Pavel R.

    2009-10-05

    This report describes recent efforts to develop glass property models that can be used to help estimate the volume of high-level waste (HLW) glass that will result from vitrification of Hanford tank waste. The compositions of acceptable and processable HLW glasses need to be optimized to minimize the waste-form volume and, hence, to save cost. A database of properties and associated compositions for simulated waste glasses was collected for developing property-composition models. This database, although not comprehensive, represents a large fraction of data on waste-glass compositions and properties that were available at the time of this report. Glass property-composition models were fit to subsets of the database for several key glass properties. These models apply to a significantly broader composition space than those previously publised. These models should be considered for interim use in calculating properties of Hanford waste glasses.

  1. Progress and future direction for the interim safe storage and disposal of Hanford high level waste (HLW)

    International Nuclear Information System (INIS)

    Wodrich, D.D.

    1996-01-01

    This paper describes the progress made at the largest environmental cleanup program in the United States. Substantial advances in methods to start interim safe storage of Hanford Site high-level wastes, waste characterization to support both safety- and disposal-related information needs, and proceeding with cost-effective disposal by the US DOE and its Hanford Site contractors, have been realized. Challenges facing the Tank Waste Remediation System Program, which is charged with the dual and parallel missions of interim safe storage and disposal of the high-level tank waste stored at the Hanford Site, are described

  2. Strategic lessons in high-level waste management planning

    International Nuclear Information System (INIS)

    Chapman, Neil

    1999-01-01

    This presentation discusses some issues in the planning and execution of high-level waste (HLW) disposal. The topics are (1) Initial considerations, (2) Issues in structuring a programme, (3) Disposal concepts, (4) Geological environments, (5) Site selection and characterisation, (6) Waste transport, (7) Performance assessment methodology and application, (8) Some key issues. The options for spent fuel management can give rise to a variety of different wastes. The quantity of waste arising will affect the volume of rock required for deposition, both with respect to rock integrity and requirements for heat dissipation. A repository must not be considered in isolation from the rest of the waste management programme. The repository development plan should be supported by a schedule of activities and related funding mechanisms, implying a long-term commitment in policy terms, and should include a corresponding legal and regulatory framework. The idea that disposed waste might be retrieved by future generations for processing under new technology is discussed. Safeguards requirements on fissile material within spent fuel or any other wastes imply indefinite control. Disposal concepts include the geological environment and the engineered barrier system within it. Site selection involves several steps: regional-scale characterisation, local characterisation, hydrological studies, etc. Key issues are retrieval vs. safeguards, optimisation of repository design, reducing long programme timescales, international collaboration

  3. Site suitability criteria for solidified high level waste repositories

    Energy Technology Data Exchange (ETDEWEB)

    Heckman, R.A.; Holdsworth, T.; Isherwood, D.; Towse, D.F.; Dayem, N.L.

    1979-04-03

    The NRC is developing a framework of regulations, criteria, and standards. Lawrence Livermore Laboratory provides broad technical support to the NRC for developing this regulatory framework, part of which involves site suitability criteria for solidified high-level wastes (SHLW). Both the regulatory framework and the technical base on which it rests have evolved in time. This document is the second report of the technical support project. It was issued as a draft working paper for a programmatic review held at LLL from August 16 to 18, 1977. It was printed and distributed solely as a briefing document on preliminary methodology and initial findings for the purpose of critical review by those in attendance. These briefing documents are being reprinted now in their original formats as UCID-series reports for the sake of the historical record. Analysis results have evolved as both the models and data base have changed. As a result, the methodology, models, and data base in this document are severely outmoded.

  4. Strategic lessons in high-level waste management planning

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, Neil

    1999-07-01

    This presentation discusses some issues in the planning and execution of high-level waste (HLW) disposal. The topics are (1) Initial considerations, (2) Issues in structuring a programme, (3) Disposal concepts, (4) Geological environments, (5) Site selection and characterisation, (6) Waste transport, (7) Performance assessment methodology and application, (8) Some key issues. The options for spent fuel management can give rise to a variety of different wastes. The quantity of waste arising will affect the volume of rock required for deposition, both with respect to rock integrity and requirements for heat dissipation. A repository must not be considered in isolation from the rest of the waste management programme. The repository development plan should be supported by a schedule of activities and related funding mechanisms, implying a long-term commitment in policy terms, and should include a corresponding legal and regulatory framework. The idea that disposed waste might be retrieved by future generations for processing under new technology is discussed. Safeguards requirements on fissile material within spent fuel or any other wastes imply indefinite control. Disposal concepts include the geological environment and the engineered barrier system within it. Site selection involves several steps: regional-scale characterisation, local characterisation, hydrological studies, etc. Key issues are retrieval vs. safeguards, optimisation of repository design, reducing long programme timescales, international collaboration.

  5. Defense High-Level Waste Leaching Mechanisms Program. Final report

    International Nuclear Information System (INIS)

    Mendel, J.E.

    1984-08-01

    The Defense High-Level Waste Leaching Mechanisms Program brought six major US laboratories together for three years of cooperative research. The participants reached a consensus that solubility of the leached glass species, particularly solubility in the altered surface layer, is the dominant factor controlling the leaching behavior of defense waste glass in a system in which the flow of leachant is constrained, as it will be in a deep geologic repository. Also, once the surface of waste glass is contacted by ground water, the kinetics of establishing solubility control are relatively rapid. The concentrations of leached species reach saturation, or steady-state concentrations, within a few months to a year at 70 to 90 0 C. Thus, reaction kinetics, which were the main subject of earlier leaching mechanisms studies, are now shown to assume much less importance. The dominance of solubility means that the leach rate is, in fact, directly proportional to ground water flow rate. Doubling the flow rate doubles the effective leach rate. This relationship is expected to obtain in most, if not all, repository situations

  6. The IFR pyroprocessing for high-level waste minimization

    International Nuclear Information System (INIS)

    Laidler, J.J.

    1993-01-01

    The process developed for the recycle of integral fast reactor (IFR) spent fuel utilizes a combination of pyrometallurgical and electrochemical methods and has been termed pyroprocessing. The process has been operated at full scale with simulated spent fuel using nonradioactive fission product elements. A comprehensive demonstration of the pyroprocessing of irradiated IFR fuel will begin later this year. Pyroprocessing involves the anodic dissolution of all the constituent elements of the IFR spent fuel and controlled electrotransport (electrorefining) to separate the actinide elements from the fission products present in the spent fuel. The process be applied to the processing of spent light water reactor (LWR) fuel as well, requiring only the addition of a reduction step to convert the LWR fuel as well, requiring only the addition of a reduction step to convert the LWR oxide fuel to metallic form and a separation step to separate uranium from the transuranic (TRU) elements. The TRU elements are then recovered by electroefining in the same manner as the actinides from the IFR high-level wastes arising from pyroprocessing are virtually free of actinides, and the volume of the wastes is minimized by the intrinsic characteristics of the processing of the processing method

  7. Safety of geological disposal of high-level waste

    International Nuclear Information System (INIS)

    Ohe, Toshiaki; Tsukamoto, Masaki

    1989-01-01

    This paper represents an analysis of barrier performance of high-level waste disposal. Advantages of a multi-barrier system in repository are checked through experiments and simulations; thermal restriction, glass-leaching, and nuclide migration in both buffer materials and surrounding rock media. The temperature distribution in a repository is calculated with TRUMP code, then the pit interval is determined according to the temperature criteria of compacted bentonite. The simulation code for glass corrosion, STRAG, is developed on the basis of the experimental findings of the JSS project in which the actual radioactive glass fabricated CEA/Marcoule was used. STRAG is then verified through agreements of the simulated and measured values. Nuclide migration in compacted bentonite is calculated by SWIFT code, and the results show the bentonite capability for retention of nuclides released from waste glass. Migration of cesium isotope in rock is also examined with the small granite core samples, of which results suggest that bulk-granite except for fractures is expected as a porous media. (author)

  8. Defense High-Level Waste Leaching Mechanisms Program. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mendel, J.E. (compiler)

    1984-08-01

    The Defense High-Level Waste Leaching Mechanisms Program brought six major US laboratories together for three years of cooperative research. The participants reached a consensus that solubility of the leached glass species, particularly solubility in the altered surface layer, is the dominant factor controlling the leaching behavior of defense waste glass in a system in which the flow of leachant is constrained, as it will be in a deep geologic repository. Also, once the surface of waste glass is contacted by ground water, the kinetics of establishing solubility control are relatively rapid. The concentrations of leached species reach saturation, or steady-state concentrations, within a few months to a year at 70 to 90/sup 0/C. Thus, reaction kinetics, which were the main subject of earlier leaching mechanisms studies, are now shown to assume much less importance. The dominance of solubility means that the leach rate is, in fact, directly proportional to ground water flow rate. Doubling the flow rate doubles the effective leach rate. This relationship is expected to obtain in most, if not all, repository situations.

  9. Evaluation of a high-level waste radiological maintenance facility

    International Nuclear Information System (INIS)

    Collins, K.J.

    1998-01-01

    The Savannah River Site''s (SRS) Defense Waste Processing Facility (DWPF) near Aiken, SC is the nation''s first and world''s largest high level waste vitrification facility. DWPF began, operations in March 1996 to process radioactive waste, consisting of a matrixed predominantly 137 Cs precipitate and a predominately 90 Sr and alpha emitting sludge, into boro-silicate glass for long term storage. Presently, DWPF is processing only sludge waste and is preparing to process a combination of sludge and precipitate waste. During precipitate operations, canister dose rates are expected to exceed 10 Sv hr -1 (1000 rem hr -1 ). In sludge-only operations, canister contact gamma dose rates are approximately 15 mSv hr -1 (1500 mrem hr -1 ). Transferable contamination levels have been greater than 10 mSv hr -1 (100 cm 2 ) -1 for beta-gamma emitters and into the millions of Bq (100 cm 2 ) -1 for the alpha emitting radionuclides. This paper presents an evaluation of the radiological maintenance areas and their ability to support radiological work

  10. Testing of high-level waste forms under repository conditions

    International Nuclear Information System (INIS)

    Mc Menamin, T.

    1989-01-01

    The workshop on testing of high-level waste forms under repository conditions was held on 17 to 21 October 1988 in Cadarache, France, and sponsored by the Commission of the European Communities (CEC), the Commissariat a l'energie atomique (CEA) and the Savannah River Laboratory (US DOE). Participants included representatives from Australia, Belgium, Denmark, France, Germany, Italy, Japan, the Netherlands, Sweden, Switzerland, The United Kingdom and the United States. The first part of the conference featured a workshop on in situ testing of simulated nuclear waste forms and proposed package components, with an emphasis on the materials interface interactions tests (MIIT). MIIT is a sevent-part programme that involves field testing of 15 glass and waste form systems supplied by seven countries, along with potential canister and overpack materials as well as geologic samples, in the salt geology at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico, USA. This effort is still in progress and these proceedings document studies and findings obtained thus far. The second part of the meeting emphasized multinational experimental studies and results derived from repository systems simulation tests (RSST), which were performed in granite, clay and salt environments

  11. Process Design Concepts for Stabilization of High Level Waste Calcine

    Energy Technology Data Exchange (ETDEWEB)

    T. R. Thomas; A. K. Herbst

    2005-06-01

    The current baseline assumption is that packaging ¡§as is¡¨ and direct disposal of high level waste (HLW) calcine in a Monitored Geologic Repository will be allowed. The fall back position is to develop a stabilized waste form for the HLW calcine, that will meet repository waste acceptance criteria currently in place, in case regulatory initiatives are unsuccessful. A decision between direct disposal or a stabilization alternative is anticipated by June 2006. The purposes of this Engineering Design File (EDF) are to provide a pre-conceptual design on three low temperature processes under development for stabilization of high level waste calcine (i.e., the grout, hydroceramic grout, and iron phosphate ceramic processes) and to support a down selection among the three candidates. The key assumptions for the pre-conceptual design assessment are that a) a waste treatment plant would operate over eight years for 200 days a year, b) a design processing rate of 3.67 m3/day or 4670 kg/day of HLW calcine would be needed, and c) the performance of waste form would remove the HLW calcine from the hazardous waste category, and d) the waste form loadings would range from about 21-25 wt% calcine. The conclusions of this EDF study are that: (a) To date, the grout formulation appears to be the best candidate stabilizer among the three being tested for HLW calcine and appears to be the easiest to mix, pour, and cure. (b) Only minor differences would exist between the process steps of the grout and hydroceramic grout stabilization processes. If temperature control of the mixer at about 80„aC is required, it would add a major level of complexity to the iron phosphate stabilization process. (c) It is too early in the development program to determine which stabilizer will produce the minimum amount of stabilized waste form for the entire HLW inventory, but the volume is assumed to be within the range of 12,250 to 14,470 m3. (d) The stacked vessel height of the hot process vessels

  12. Idaho National Engineering Laboratory High-Level Waste Roadmap

    International Nuclear Information System (INIS)

    1993-08-01

    The Idaho National Engineering Laboratory (INEL) High-Level Waste (HLW) Roadmap takes a strategic look at the entire HLW life-cycle starting with generation, through interim storage, treatment and processing, transportation, and on to final disposal. The roadmap is an issue-based planning approach that compares ''where we are now'' to ''where we want and need to be.'' The INEL has been effectively managing HLW for the last 30 years. Calcining operations are continuing to turn liquid HLW into a more manageable form. Although this document recognizes problems concerning HLW at the INEL, there is no imminent risk to the public or environment. By analyzing the INEL current business operations, pertinent laws and regulations, and committed milestones, the INEL HLW Roadmap has identified eight key issues existing at the INEL that must be resolved in order to reach long-term objectives. These issues are as follows: A. The US Department of Energy (DOE) needs a consistent policy for HLW generation, handling, treatment, storage, and disposal. B. The capability for final disposal of HLW does not exist. C. Adequate processes have not been developed or implemented for immobilization and disposal of INEL HLW. D. HLW storage at the INEL is not adequate in terms of capacity and regulatory requirements. E. Waste streams are generated with limited consideration for waste minimization. F. HLW is not adequately characterized for disposal nor, in some cases, for storage. G. Research and development of all process options for INEL HLW treatment and disposal are not being adequately pursued due to resource limitations. H. HLW transportation methods are not selected or implemented. A root-cause analysis uncovered the underlying causes of each of these issues

  13. High-level waste issues and resolutions document

    International Nuclear Information System (INIS)

    1994-05-01

    The High-Level Waste (HLW) Issues and Resolutions Document recognizes US Department of Energy (DOE) complex-wide HLW issues and offers potential corrective actions for resolving these issues. Westinghouse Management and Operations (M ampersand O) Contractors are effectively managing HLW for the Department of Energy at four sites: Idaho National Engineering Laboratory (INEL), Savannah River Site (SRS), West Valley Demonstration Project (WVDP), and Hanford Reservation. Each site is at varying stages of processing HLW into a more manageable form. This HLW Issues and Resolutions Document identifies five primary issues that must be resolved in order to reach the long-term objective of HLW repository disposal. As the current M ampersand O contractor at DOE's most difficult waste problem sites, Westinghouse recognizes that they have the responsibility to help solve some of the complexes' HLW problems in a cost effective manner by encouraging the M ampersand Os to work together by sharing expertise, eliminating duplicate efforts, and sharing best practices. Pending an action plan, Westinghouse M ampersand Os will take the initiative on those corrective actions identified as the responsibility of an M ampersand O. This document captures issues important to the management of HLW. The proposed resolutions contained within this document set the framework for the M ampersand Os and DOE work cooperatively to develop an action plan to solve some of the major complex-wide problems. Dialogue will continue between the M ampersand Os, DOE, and other regulatory agencies to work jointly toward the goal of storing, treating, and immobilizing HLW for disposal in an environmentally sound, safe, and cost effective manner

  14. Computationally based methodology for reengineering the high-level waste planning process at SRS

    International Nuclear Information System (INIS)

    Paul, P.K.; Gregory, M.V.; Wells, M.N.

    1997-01-01

    The Savannah River Site (SRS) has started processing its legacy of 34 million gallons of high-level radioactive waste into its final disposable form. The SRS high-level waste (HLW) complex consists of 51 waste storage tanks, 3 evaporators, 6 waste treatment operations, and 2 waste disposal facilities. It is estimated that processing wastes to clean up all tanks will take 30+ yr of operation. Integrating all the highly interactive facility operations through the entire life cycle in an optimal fashion-while meeting all the budgetary, regulatory, and operational constraints and priorities-is a complex and challenging planning task. The waste complex operating plan for the entire time span is periodically published as an SRS report. A computationally based integrated methodology has been developed that has streamlined the planning process while showing how to run the operations at economically and operationally optimal conditions. The integrated computational model replaced a host of disconnected spreadsheet calculations and the analysts' trial-and-error solutions using various scenario choices. This paper presents the important features of the integrated computational methodology and highlights the parameters that are core components of the planning process

  15. Final report, Task 4: options for on-site management of Nuclear Fuel Services, Inc. high level waste

    International Nuclear Information System (INIS)

    1978-01-01

    Two on-site management options for handling the NFS high-level waste were analyzed: in-tank cement solidification and perpetual tank storage of the liquid waste. The cost of converting the 8D4 plus 8D2 waste to a cementitious solid, including mixing, grout preparation, and transfer to tank 8D1 would require $3,651,000; the cost of cooling the solidified solid for 15 years, plus the cost of filling the rest of the tank space and annulus with grout, plus the cost of minimum surveillance are $10,002,000. Modification of tank 8D2 would be required; prior to transfer of the waste, tank 8D1 would also be modified for cooling of the grout mass. Estimated costs of perpetual tank storage (replacing the existing neutralized waste tank after 10 years, then transferring contents at 50-y intervals for 1000 y, with replacement of ventilation system and auxiliaries at 30-y intervals) would require a sinking fund of $11,039,000. The acidic 8D4 waste would be transferred at 50-y intervals. The sinking fund requirements are sensitive to the difference between the interest rate and the escalation rate, and also to the time assumed from present to the first tank replacement

  16. Independent Assessment of the Savannah River Site High-Level Waste Salt Disposition Alternatives Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    J. T. Case (DOE-ID); M. L. Renfro (INEEL)

    1998-12-01

    This report presents the results of the Independent Project Evaluation (IPE) Team assessment of the Westinghouse Savannah River Company High-Level Waste Salt Disposition Systems Engineering (SE) Team's deliberations, evaluations, and selections. The Westinghouse Savannah River Company concluded in early 1998 that production goals and safety requirements for processing SRS HLW salt to remove Cs-137 could not be met in the existing In-Tank Precipitation Facility as currently configured for precipitation of cesium tetraphenylborate. The SE Team was chartered to evaluate and recommend an alternative(s) for processing the existing HLW salt to remove Cs-137. To replace the In-Tank Precipitation process, the Savannah River Site HLW Salt Disposition SE Team downselected (October 1998) 140 candidate separation technologies to two alternatives: Small-Tank Tetraphenylborate (TPB) Precipitation (primary alternative) and Crystalline Silicotitanate (CST) Nonelutable Ion Exchange (backup alternative). The IPE Team, commissioned by the Department of Energy, concurs that both alternatives are technically feasible and should meet all salt disposition requirements. But the IPE Team judges that the SE Team's qualitative criteria and judgments used in their downselection to a primary and a backup alternative do not clearly discriminate between the two alternatives. To properly choose between Small-Tank TPB and CST Ion Exchange for the primary alternative, the IPE Team suggests the following path forward: Complete all essential R and D activities for both alternatives and formulate an appropriate set of quantitative decision criteria that will be rigorously applied at the end of the R and D activities. Concurrent conceptual design activities should be limited to common elements of the alternatives.

  17. High-level waste repository-induced effects

    International Nuclear Information System (INIS)

    Leupin, O.X.; Marschall, P.; Johnson, L.; Cloet, V.; Schneider, J.; Smith, P.; Savage, D.; Senger, R.

    2016-10-01

    This status report aims at describing and assessing the interactions of the radioactive waste emplaced in a high-level waste (HLW) repository with the engineered materials and the Opalinus Clay host rock. The Opalinus Clay has a thickness of about 100 m in the proposed siting regions. Among other things the results are used to steer the RD and D programme of NAGRA. The repository-induced effects considered in this report are of the following broad types: - Thermal effects: i.e. effects on the host rock and engineered barriers arising principally from the heat generated by the waste. - Rock-mechanical effects: i.e. effects arising from the mechanical disturbance to the rock caused by the excavation of the emplacement rooms and other underground structures. - Hydraulic and gas-related effects: i.e. the effects of repository resaturation and of gas generation, e.g. due to the corrosion of metals within the repository, on the host rock and engineered barriers. - Chemical effects: i.e. chemical interactions between the waste, the engineered materials and the host rock, with a focus on chemical effects of the waste and engineered materials on the host rock. The assessment of the repository-induced effects shows that detrimental chemical and mechanical impacts are largely confined to the rock immediately adjacent to the excavations, thermal impacts are controllable by limiting the heat load and gas effects are limited by ensuring acceptably low gas production rates and by the natural tendency of the gas to escape along the excavations and the excavation damaged zone (EDZ) rather than through the undisturbed rock. Specific measures that are part of the current reference design are discussed in relation to their significance with respect to repository-induced effects. The SF/HLW emplacement rooms (emplacement drifts) are designed, constructed, operated and finally backfilled in such a way that formation of excavation damaged zones is limited. Specifically this is achieved

  18. Interaction of cementitious materials with high-level waste

    International Nuclear Information System (INIS)

    Lemmens, Karel; Cachoir, Christelle; Ferrand, Karine; Mennecart, Thierry; Gielen, Ben; Vercauter, Regina

    2012-01-01

    Document available in abstract form only: Since a few years, the Belgian agency for radioactive waste (ONDRAF/NIRAS) has selected the Supercontainer design with an Ordinary Portland Cement (OPC) buffer as the reference design for geological disposal of High-Level Waste (HLW) and Spent Fuel (SF) in the Boom Clay formation. The Boom Clay beneath the Mol-Dessel nuclear zone is a reference methodological site for supporting R and D. Compared to the previous bentonite based reference design, described in detail in the final SAFIR 2 report, the supercontainer will provide a highly alkaline chemical environment allowing the passivation of the surface of the overpack and the inhibition of its corrosion. The Supercontainer will contribute to the containment of radionuclides, but it will also have an effect on the retardation of radionuclide release from the waste and it will retard the migration of the released radionuclides. In the Supercontainer design, the canisters of HLW or SF will be enclosed by a 30 mm thick carbon steel overpack and a concrete buffer about 700 mm thick. The overpack will prevent contact with the (cementitious) pore water during the thermal phase. On the other hand, once the overpack will be locally perforated, the high pH of the incoming water may have an impact on the lifetime of the vitrified waste or spent fuel. The behaviour of these waste forms in disposal conditions has been studied for several decades, but the vast majority of published data is related to the interaction with backfill or host rock materials at near-neutral pH. Very few studies have been reported for alkaline media, at pH >11. Hence, a research programme including new experiments, was started at the Belgian Nuclear Research Centre (SCK.CEN) and at INE (FZK) to assess the rate at which the radionuclides are released by the vitrified waste and spent fuel in such an environment. The presence of concrete will have an impact on the behaviour of the vitrified HLW and spent fuel. For

  19. High level waste facilities - Continuing operation or orderly shutdown

    International Nuclear Information System (INIS)

    Decker, L.A.

    1998-04-01

    Two options for Environmental Impact Statement No action alternatives describe operation of the radioactive liquid waste facilities at the Idaho Chemical Processing Plant at the Idaho National Engineering and Environmental Laboratory. The first alternative describes continued operation of all facilities as planned and budgeted through 2020. Institutional control for 100 years would follow shutdown of operational facilities. Alternatively, the facilities would be shut down in an orderly fashion without completing planned activities. The facilities and associated operations are described. Remaining sodium bearing liquid waste will be converted to solid calcine in the New Waste Calcining Facility (NWCF) or will be left in the waste tanks. The calcine solids will be stored in the existing Calcine Solids Storage Facilities (CSSF). Regulatory and cost impacts are discussed

  20. TWRS retrieval and disposal mission. Immobilized high-level waste storage plan

    International Nuclear Information System (INIS)

    Calmus, R.B.

    1998-01-01

    This project plan has a two fold purpose. First, it provides a plan specific to the Hanford Tank Waste Remediation System (TWRS) Immobilized High-Level Waste (EMW) Storage Subproject for the Washington State Department of Ecology (Ecology) that meets the requirements of Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) milestone M-90-01 (Ecology et al. 1996) and is consistent with the project plan content guidelines found in Section 11.5 of the Tri-Party Agreement action plan. Second, it provides an upper tier document that can be used as the basis for future subproject line item construction management plans. The planning elements for the construction management plans are derived from applicable U.S. Department of Energy (DOE) planning guidance documents (DOE Orders 4700.1 (DOE 1992a) and 430.1 (DOE 1995)). The format and content of this project plan are designed to accommodate the plan's dual purpose. A cross-check matrix is provided in Appendix A to explain where in the plan project planning elements required by Section 11.5 of the Tri-Party Agreement are addressed

  1. Development of Simulants to Support Mixing Tests for High Level Waste and Low Activity Waste

    International Nuclear Information System (INIS)

    EIBLING, RUSSELLE.

    2004-01-01

    The objectives of this study were to develop two different types of simulants to support vendor agitator design studies and mixing studies. The initial simulant development task was to develop rheologically-bounding physical simulants and the final portion was to develop a nominal chemical simulant which is designed to match, as closely as possible, the actual sludge from a tank. The physical simulants to be developed included a lower and upper rheologically bounded: pretreated low activity waste (LAW) physical simulant; LAW melter feed physical simulant; pretreated high level waste (HLW) physical simulant; HLW melter feed physical simulant. The nominal chemical simulant, hereafter referred to as the HLW Precipitated Hydroxide simulant, is designed to represent the chemical/physical composition of the actual washed and leached sludge sample. The objective was to produce a simulant which matches not only the chemical composition but also the physical properties of the actual waste sample. The HLW Precipitated Hydroxide simulant could then be used for mixing tests to validate mixing, homogeneity and representative sampling and transferring issues. The HLW Precipitated Hydroxide simulant may also be used for integrated nonradioactive testing of the WTP prior to radioactive operation

  2. A dynamic simulation model of the Savannah River Site high level waste complex

    International Nuclear Information System (INIS)

    Gregory, M.V.; Aull, J.E.; Dimenna, R.A.

    1994-01-01

    A detailed, dynamic simulation entire high level radioactive waste complex at the Savannah River Site has been developed using SPEEDUP(tm) software. The model represents mass transfer, evaporation, precipitation, sludge washing, effluent treatment, and vitrification unit operation processes through the solution of 7800 coupled differential and algebraic equations. Twenty-seven discrete chemical constituents are tracked through the unit operations. The simultaneous simultaneous simulation of concurrent batch and continuous processes is achieved by several novel, customized SPEEDUP(tm) algorithms. Due to the model's computational burden, a high-end work station is required: simulation of a years operation of the complex requires approximately three CPU hours on an IBM RS/6000 Model 590 processor. The model will be used to develop optimal high level waste (HLW) processing strategies over a thirty year time horizon. It will be employed to better understand the dynamic inter-relationships between different HLW unit operations, and to suggest strategies that will maximize available working tank space during the early years of operation and minimize overall waste processing cost over the long-term history of the complex. Model validation runs are currently underway with comparisons against actual plant operating data providing an excellent match

  3. Systems costs for disposal of Savannah River high-level waste sludge and salt

    International Nuclear Information System (INIS)

    McDonell, W.R.; Goodlett, C.B.

    1984-01-01

    A systems cost model has been developed to support disposal of defense high-level waste sludge and salt generated at the Savannah River Plant. Waste processing activities covered by the model include decontamination of the salt by a precipitation process in the waste storage tanks, incorporation of the sludge and radionuclides removed from the salt into glass in the Defense Waste Processing Facility (DWPF), and, after interim storage, final disposal of the DWPF glass waste canisters in a federal geologic repository. Total costs for processing of waste generated to the year 2000 are estimated to be about $2.9 billion (1984 dollars); incremental unit costs for DWPF and repository disposal activities range from $120,000 to $170,000 per canister depending on DWPF processing schedules. In a representative evaluation of process alternatives, the model is used to demonstrate cost effectiveness of adjustments in the frit content of the waste glass to reduce impacts of wastes generated by the salt decontamination operations. 13 references, 8 tables

  4. Aluminosilicate Formation in High Level Waste Evaporators: A Mechanism for Uranium Accumulation

    International Nuclear Information System (INIS)

    Wilmarth, W.R.

    2002-01-01

    High level waste Evaporators at the Savannah River Site (SRS) process radioactive waste to concentrate supernate and thus conserve tank space. In June of 1997, difficulty in evaporator operation was initially observed. This operational difficulty evidenced itself as a plugging of the evaporator's gravity drain line (GDL). The material blocking the GDL was determined to be a sodium aluminosilicate. Following a mechanical cleaning of the GDL, the evaporator was returned to service until October 1999. At this time massive deposits were discovered in the evaporator pot. As a result of the changes in evaporator chemistry and the resulting formation of aluminosilicate deposits in the evaporator, a comprehensive research and development program has been undertaken. This program is underway in order to assist in understanding the new evaporator chemistry and gain insight into the deposition phenomena. Key results from testing in FY01 have demonstrated that the chemistry of the evaporator feed favors aluminosilicate formation. Both the reaction kinetics and particle growth of the aluminosilicate material under SRS evaporator conditions has been demonstrated to occur within the residence times utilized in the SRS evaporator operation. Batch and continuous-flow experiments at known levels of supersaturation have shown a significant correlation between the deposition of aluminosilicates and mixing intensity in the vessel. Advances in thermodynamic modeling of the evaporator chemistry have been accomplished. The resulting thermodynamic model has been related to the operational history of the evaporator, is currently assisting in feed selection, and could potentially assist in expanding the operating envelopes technical baselines for evaporator operation

  5. NOx AND HETEROGENEITY EFFECTS IN HIGH LEVEL WASTE (HLW)

    International Nuclear Information System (INIS)

    Meisel, Dan; Camaioni, Donald M.; Orlando, Thom

    2000-01-01

    We summarize contributions from our EMSP supported research to several field operations of the Office of Environmental Management (EM). In particular we emphasize its impact on safety programs at the Hanford and other EM sites where storage, maintenance and handling of HLW is a major mission. In recent years we were engaged in coordinated efforts to understand the chemistry initiated by radiation in HLW. Three projects of the EMSP (''The NOx System in Nuclear Waste,'' ''Mechanisms and Kinetics of Organic Aging in High Level Nuclear Wastes, D. Camaioni--PI'' and ''Interfacial Radiolysis Effects in Tanks Waste, T. Orlando--PI'') were involved in that effort, which included a team at Argonne, later moved to the University of Notre Dame, and two teams at the Pacific Northwest National Laboratory. Much effort was invested in integrating the results of the scientific studies into the engineering operations via coordination meetings and participation in various stages of the resolution of some of the outstanding safety issues at the sites. However, in this Abstract we summarize the effort at Notre Dame

  6. Disposition of actinides released from high-level waste glass

    International Nuclear Information System (INIS)

    Ebert, W.L.; Bates, J.K.; Buck, E.C.; Gong, M.; Wolf, S.F.

    1994-01-01

    A series of static leach tests was conducted using glasses developed for vitrifying tank wastes at the Savannah River Site to monitor the disposition of actinide elements upon corrosion of the glasses. In these tests, glasses produced from SRL 131 and SRL 202 frits were corroded at 90 degrees C in a tuff groundwater. Tests were conducted using crushed glass at different glass surface area-to-solution volume (S/V) ratios to assess the effect of the S/V on the solution chemistry, the corrosion of the glass, and the disposition of actinide elements. Observations regarding the effects of the S/V on the solution chemistry and the corrosion of the glass matrix have been reported previously. This paper highlights the solution analyses performed to assess how the S/V used in a static leach test affects the disposition of actinide elements between fractions that are suspended or dissolved in the solution, and retained by the altered glass or other materials

  7. Effects of aqueous-soluble organic compounds on the removal of selected radionuclides from high-level waste part I: Distribution of Sr, Cs, and Tc onto 18 absorbers from an irradiated, organic-containing leachate simulant for Hanford Tank 101-SY

    Energy Technology Data Exchange (ETDEWEB)

    Marsh, S.F. [Sandia National Labs., Albuquerque, NM (United States); Svitra, Z.V.; Bowen, S.M. [Los Alamos National Lab., NM (United States)

    1995-01-01

    Many of the radioactive waste storage tanks at U.S. Department of Energy facilities contain organic compounds that have been degraded by radiolysis and chemical reactions. In this investigation, we measured the effect of some aqueous-soluble organic compounds on the sorption of strontium, cesium, and technetium onto 18 absorbers that offer high sorption of strontium from organic-free solutions. For our test solution we used a leachate from a simulated slurry for Hanford Tank 101-SY that initially contained ethylenediaminetetraacetic acid (EDTA) and then was gamma-irradiated to 34 Mrads. We measured distribution coefficients (Kds) for each element/absorber combination for dynamic contact periods of 30 min, 2 h, and 6 h to obtain information about sorption kinetics. To facilitate comparisons, we include Kd values for these same element/absorber combinations from three organic-free simulant solutions. The Kd values for strontium sorption from the simulant that contained the degraded organics usually decreased by large factors, whereas the Kd values for cesium and technetium sorption were relatively unaffected.

  8. Effects of soluble organic complexants and their degradation products on the removal of selected radionuclides from high-level waste. Part II: Distributions of Sr, Cs, Tc, and Am onto 32 absorbers from four variations of Hanford tank 101-SY simulant solution

    Energy Technology Data Exchange (ETDEWEB)

    Marsh, S.F. [Sandia National Labs., Albuquerque, NM (United States); Svitra, Z.V.; Bowen, S.M. [Los Alamos National Lab., NM (United States)

    1995-04-01

    Many of the radioactive waste storage tanks at U.S. Department of Energy facilities contain organic compounds that have been degraded by radiolysis and chemical reactions during decades of storage. In this second part of our three-part investigation of the effects of soluble organic complexants and their degradation products, we measured the sorption of strontium, cesium, technetium, and americium onto 32 absorbers that offer high sorption of these elements in the absence of organic complexants. The four solutions tested were (1) a simulant for a 3:1 dilution of Hanford Tank 101-SY contents that initially contained ethylenediaminetetraacetic acid (EDTA), (2) this simulant after gamma-irradiation to 34 Mrads, (3) the unirradiated simulant after treatment with a hydrothermal organic-destruction process, and (4) the irradiated simulant after hydrothermal processing. For each of 512 element/absorber/solution combinations, we measured distribution coefficients (Kds) twice for each period for dynamic contact periods of 30 min, 2 h, and 6 h to obtain information about sorption kinetics. On the basis of our 3,072 measured Kd values, the sorption of strontium and americium is significantly decreased by the organic components of the simulant solutions, whereas the sorption of cesium and technetium appears unaffected by the organic components of the simulant solutions.

  9. Preliminary estimates of cost savings for defense high level waste vitrification options

    International Nuclear Information System (INIS)

    Merrill, R.A.; Chapman, C.C.

    1993-09-01

    The potential for realizing cost savings in the disposal of defense high-level waste through process and design modificatins has been considered. Proposed modifications range from simple changes in the canister design to development of an advanced melter capable of processing glass with a higher waste loading. Preliminary calculations estimate the total disposal cost (not including capital or operating costs) for defense high-level waste to be about $7.9 billion dollars for the reference conditions described in this paper, while projected savings resulting from the proposed process and design changes could reduce the disposal cost of defense high-level waste by up to $5.2 billion

  10. High-Level Waste Salt Disposition Systems Engineering Team Final Report, Volumes I, II, and III

    International Nuclear Information System (INIS)

    Piccolo, S.F.

    1999-01-01

    This report describes the process used and results obtained by the High Level Waste Salt Disposition Systems Engineering Team to select a primary and backup alternative salt disposition method for the Savannah River Site

  11. High level waste containing granules coated and embedded in metal as an alternative to HLW glasses

    International Nuclear Information System (INIS)

    Neumann, W.

    1980-01-01

    Simulated high level waste containing granules were overcoated with pyrocarbon or nickel respectively. The coatings were performed by the use of chemical vapour deposition in a fluidized bed. The coated granules were embedded in an aluminium-silicon-alloy to improve the dissipation of radiation induced heat. The metal-granules-composites obtained were of improved product stability related to the high level waste containing glasses. (orig.) [de

  12. The Canadian program for management of spent fuel and high level wastes

    International Nuclear Information System (INIS)

    Barnes, R.W.; Mayman, S.A.

    A brief history and description of the nuclear power program in Canada is given. Schedules and programs are described for storing spent fuel in station fuel bays, centralized water pool storage facilities, concrete canisters, convection vaults, and rock or salt formations. High-level wastes will be retrievable initially, therefore the focus is on storage in mined cavities. The methods developed for high-level waste storage/disposal will ideally be flexible enough to accommodate spent fuel. (E.C.B.)

  13. Sandia studies of high-level waste canisters and overpacks applicable for a salt repository

    International Nuclear Information System (INIS)

    Molecke, M.A.; Schaefer, D.W.; Glass, R.S.; Ruppen, J.A.

    1982-01-01

    An experimental program to develop candidate materials for use as high-level waste (HLW) overpacks or canisters in a salt repository has been in progress at Sandia National Laboratories since 1976. The main objective of this program is to provide a waste package barrier having a long lifetime in the chemical and physical environment of a repository. This paper summarizes the recent corrosion and metallurgical study results for the prime overpack material, TiCode-12, in the areas of uniform corrosion (extremely low rate and extent); local attack, e.g., pits and crevices (none were found); stress corrosion cracking susceptibility (no significant changes in macroscopic tensile properties were detected); hydrogen sorption-embrittlement effects; effects of gamma irradiation in solution; and sensitization effects (testing is still in process in the last three areas). Previous candidate screening analyses on other alloys and recent work on alternate overpack alloys are reviewed. All phases of these interrelated laboratory, hot-cell, and field experimental studies are described. 16 references, 8 figures, 4 tables

  14. A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests

    Energy Technology Data Exchange (ETDEWEB)

    Thien, Mike G. [Washington River Protection Solutions, LLC, Richland, WA (United States); Barnes, Steve M. [URS, Richland, WA (United States)

    2013-01-17

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described.

  15. A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests - 13342

    Energy Technology Data Exchange (ETDEWEB)

    Thien, Mike G. [Washington River Protection Solutions, LLC, P.O Box 850, Richland WA, 99352 (United States); Barnes, Steve M. [Waste Treatment Plant, 2435 Stevens Center Place, Richland WA 99354 (United States)

    2013-07-01

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described. (authors)

  16. A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests

    International Nuclear Information System (INIS)

    Thien, Mike G.; Barnes, Steve M.

    2013-01-01

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described

  17. RECENT PROCESS AND EQUIPMENT IMPROVEMENTS TO INCREASE HIGH LEVEL WASTE THROUGHPUT AT THE DEFENSE WASTE PROCESSING FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Odriscoll, R; Allan Barnes, A; Jim Coleman, J; Timothy Glover, T; Robert Hopkins, R; Dan Iverson, D; Jeff Leita, J

    2008-01-15

    The Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF) began stabilizing high level waste (HLW) in a glass matrix in 1996. Over the past few years, there have been several process and equipment improvements at the DWPF to increase the rate at which the high level waste can be stabilized. These improvements have either directly increased waste processing rates or have desensitized the process to upsets, thereby minimizing downtime and increasing production. Improvements due to optimization of waste throughput with increased HLW loading of the glass resulted in a 6% waste throughput increase based upon operational efficiencies. Improvements in canister production include the pour spout heated bellows liner (5%), glass surge (siphon) protection software (2%), melter feed pump software logic change to prevent spurious interlocks of the feed pump with subsequent dilution of feed stock (2%) and optimization of the steam atomized scrubber (SAS) operation to minimize downtime (3%) for a total increase in canister production of 12%. A number of process recovery efforts have allowed continued operation. These include the off gas system pluggage and restoration, slurry mix evaporator (SME) tank repair and replacement, remote cleaning of melter top head center nozzle, remote melter internal inspection, SAS pump J-Tube recovery, inadvertent pour scenario resolutions, dome heater transformer bus bar cooling water leak repair and new Infra-red camera for determination of glass height in the canister are discussed.

  18. A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests - 13342

    International Nuclear Information System (INIS)

    Thien, Mike G.; Barnes, Steve M.

    2013-01-01

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described. (authors)

  19. Chemical analysis of simulated high level waste glasses to support stage III sulfate solubility modeling

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-03-17

    The U.S. Department of Energy (DOE), Office of Environmental Management (EM) is sponsoring an international, collaborative project to develop a fundamental model for sulfate solubility in nuclear waste glass. The solubility of sulfate has a significant impact on the achievable waste loading for nuclear waste forms within the DOE complex. These wastes can contain relatively high concentrations of sulfate, which has low solubility in borosilicate glass. This is a significant issue for low-activity waste (LAW) glass and is projected to have a major impact on the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Sulfate solubility has also been a limiting factor for recent high level waste (HLW) sludge processed at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). The low solubility of sulfate in glass, along with melter and off-gas corrosion constraints, dictate that the waste be blended with lower sulfate concentration waste sources or washed to remove sulfate prior to vitrification. The development of enhanced borosilicate glass compositions with improved sulfate solubility will allow for higher waste loadings and accelerate mission completion.The objective of the current scope being pursued by SHU is to mature the sulfate solubility model to the point where it can be used to guide glass composition development for DWPF and WTP, allowing for enhanced waste loadings and waste throughput at these facilities. A series of targeted glass compositions was selected to resolve data gaps in the model and is identified as Stage III. SHU fabricated these glasses and sent samples to SRNL for chemical composition analysis. SHU will use the resulting data to enhance the sulfate solubility model and resolve any deficiencies. In this report, SRNL provides chemical analyses for the Stage III, simulated HLW glasses fabricated by SHU in support of the sulfate solubility model development.

  20. MELT RATE ENHANCEMENT FOR HIGH ALUMINUM HLW (HIGH LEVEL WASTE) GLASS FORMULATION. FINAL REPORT 08R1360-1

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Kot, W.; Pegg, I.L.; Joseph, I.; Bardakci, T.; Gan, H.; Gong, W.; Chaudhuri, M.

    2010-01-01

    This report describes the development and testing of new glass formulations for high aluminum waste streams that achieve high waste loadings while maintaining high processing rates. The testing was based on the compositions of Hanford High Level Waste (HLW) with limiting concentrations of aluminum specified by the Office of River Protection (ORP). The testing identified glass formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts and small scale melt rate screening tests. The results were used to select compositions for subsequent testing in a DuraMelter 100 (DM100) system. These tests were used to determine processing rates for the selected formulations as well as to examine the effects of increased glass processing temperature, and the form of aluminum in the waste simulant. Finally, one of the formulations was selected for large-scale confirmatory testing on the HLW Pilot Melter (DM1200), which is a one third scale prototype of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW melter and off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy (DOE) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same high-aluminum waste composition used in the present work and other Hanford HLW compositions. The scope of this study was outlined in a Test Plan that was prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the WTP is about 13,500 (equivalent to 40,500 MT glass). This estimate is based upon the inventory of the tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form

  1. MELT RATE ENHANCEMENT FOR HIGH ALUMINUM HLW (HIGH LEVEL WASTE) GLASS FORMULATION FINAL REPORT 08R1360-1

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; KOT W; PEGG IL; JOSEPH I; BARDAKCI T; GAN H; GONG W; CHAUDHURI M

    2010-01-04

    This report describes the development and testing of new glass formulations for high aluminum waste streams that achieve high waste loadings while maintaining high processing rates. The testing was based on the compositions of Hanford High Level Waste (HLW) with limiting concentrations of aluminum specified by the Office of River Protection (ORP). The testing identified glass formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts and small scale melt rate screening tests. The results were used to select compositions for subsequent testing in a DuraMelter 100 (DM100) system. These tests were used to determine processing rates for the selected formulations as well as to examine the effects of increased glass processing temperature, and the form of aluminum in the waste simulant. Finally, one of the formulations was selected for large-scale confirmatory testing on the HLW Pilot Melter (DM1200), which is a one third scale prototype of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW melter and off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy (DOE) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same high-aluminum waste composition used in the present work and other Hanford HLW compositions. The scope of this study was outlined in a Test Plan that was prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the WTP is about 13,500 (equivalent to 40,500 MT glass). This estimate is based upon the inventory of the tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form

  2. Evaluation of high-level waste vitrification feed preparation chemistry for an NCAW simulant, FY 1994: Alternate flowsheets (DRAFT)

    International Nuclear Information System (INIS)

    Smith, H.D.; Merz, M.D.; Wiemers, K.D.; Smith, G.L.

    1996-02-01

    High-level radioactive waste stored in tanks at the U.S. Department of Energy's (DOE's) Hanford Site will be pretreated to concentrate radioactive constituents and fed to the vitrification plant A flowsheet for feed preparation within the vitrification plant (based on the Hanford Waste Vitrification Plant (HWVP) design) called for HCOOH addition during the feed preparation step to adjust rheology and glass redox conditions. However, the potential for generating H 2 and NH 3 during treatment of high-level waste (HLW) with HCOOH was identified at Pacific Northwest Laboratory (PNL). Studies at the University of Georgia, under contract with Savannah River Technology Center (SRTC) and PNL, have verified the catalytic role of noble metals (Pd, Rh, Ru), present in the waste, in the generation of H 2 and NH 3 . Both laboratory-scale and pilot-scale studies at SRTC have documented the H 2 and NH 3 generation phenomenal Because H 2 and NH 3 may create hazardous conditions in the vessel vapor space and offgas system of a vitrification plant, reducing the H 2 generation rate and the NH 3 generation to the lowest possible levels consistent with desired melter feed characteristics is important. The Fiscal Year 1993 and 1994 studies were conducted with simulated (non-radioactive), pre-treated neutralized current acid waste (NCAW). Neutralized current acid waste is a high-level waste originating from the plutonium/uranium extraction (PUREX) plant that has been partially denitrated with sugar, neutralized with NaOH, and is presently stored in double-shell tanks. The non-radioactive simulant used for the present study includes all of the trace components found in the waste, or substitutes a chemically similar element for radioactive or very toxic species. The composition and simulant preparation steps were chosen to best simulate the chemical processing characteristics of the actual waste

  3. Performance of high level waste forms and engineered barriers under repository conditions

    International Nuclear Information System (INIS)

    1991-02-01

    The IAEA initiated in 1977 a co-ordinated research programme on the ''Evaluation of Solidified High-Level Waste Forms'' which was terminated in 1983. As there was a continuing need for international collaboration in research on solidified high-level waste form and spent fuel, the IAEA initiated a new programme in 1984. The new programme, besides including spent fuel and SYNROC, also placed greater emphasis on the effect of the engineered barriers of future repositories on the properties of the waste form. These engineered barriers included containers, overpacks, buffer and backfill materials etc. as components of the ''near-field'' of the repository. The Co-ordinated Research Programme on the Performance of High-Level Waste Forms and Engineered Barriers Under Repository Conditions had the objectives of promoting the exchange of information on the experience gained by different Member States in experimental performance data and technical model evaluation of solidified high level waste forms, components of the waste package and the complete waste management system under conditions relevant to final repository disposal. The programme includes studies on both irradiated spent fuel and glass and ceramic forms as the final solidified waste forms. The following topics were discussed: Leaching of vitrified high-level wastes, modelling of glass behaviour in clay, salt and granite repositories, environmental impacts of radionuclide release, synroc use for high--level waste solidification, leachate-rock interactions, spent fuel disposal in deep geologic repositories and radionuclide release mechanisms from various fuel types, radiolysis and selective leaching correlated with matrix alteration. Refs, figs and tabs

  4. Tank waste remediation system immobilized high-level waste storage project configuration management implementation plan

    International Nuclear Information System (INIS)

    Burgard, K.G.; Schaus, P.S.; Rossi, H.

    1998-01-01

    This Configuration Management Implementation Plan was developed to assist in the management of systems, structures, and components, to facilitate the effective control and statusing of changes to systems, structures, and components; and to ensure technical consistency between design, performance, and operational requirements. Its purpose is to describe the approach Project W-464 will take in implementing a configuration management control, to determine the rigor of control, and to identify the mechanisms for imposing that control.This Configuration Management Implementation Plan was developed to assist in the management of systems, structures, and components, to facilitate the effective control and statusing of changes to systems, structures, and components; and to ensure technical consistency between design, performance, and operational requirements. Its purpose is to describe the approach Project W-464 will take in implementing a configuration management control, to determine the rigor of control, and to identify the mechanisms for imposing that control

  5. High level waste storage tanks 242-A evaporator S/RID phase II assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Biebesheimer, E.

    1996-09-27

    This document, the Standards/Requirements Identification Document (S/RID) Phase 2 Assessment Report for the subject facility, represents the results of a Performance Assessment to determine whether procedures containing S/RID requirements are fully implemented by field personnel in the field. It contains a summary report and three attachments; an assessment schedule, performance objectives, and assessments for selected functional areas.

  6. High level waste storage tanks 242-A evaporator standards/requirement identification document

    International Nuclear Information System (INIS)

    Biebesheimer, E.

    1996-01-01

    This document, the Standards/Requirements Identification Document (S/RIDS) for the subject facility, represents the necessary and sufficient requirements to provide an adequate level of protection of the worker, public health and safety, and the environment. It lists those source documents from which requirements were extracted, and those requirements documents considered, but from which no requirements where taken. Documents considered as source documents included State and Federal Regulations, DOE Orders, and DOE Standards

  7. Alternative concepts for treatment and disposal of Hanford site high-level waste in tanks

    Energy Technology Data Exchange (ETDEWEB)

    Claghorn, R.D.; Powell, W.J.

    1994-12-01

    Some innovative approaches have recently been proposed that may have significant schedule, cost, or environmental advantages which could improve the current HLW program strategy. Three general categories of alternative concepts are now under consideration: (1) process/product alternatives, (2) facility layout options, and (3) contracting strategies. This report compares the alternate approaches to the current program baseline to illustrate their potential significance and to identify the risks associated with each approach.

  8. Using depleted uranium to shield vitrified high-level waste packages

    International Nuclear Information System (INIS)

    Yoshimura, H.R.; Gildea, P.D.; Bernard, E.A.

    1995-01-01

    The underlying report for this paper evaluates options for using depleted uranium as shielding materials for transport systems for disposal of vitrified high-level waste (VHLW). In addition, economic analyses are presented to compare costs associated with these options to costs, associated with existing and proposed storage, transport, and diposal capabilities. A more detailed evaluation is provided elsewhere. (Yoshimura et al. 1995.)

  9. Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)

    Energy Technology Data Exchange (ETDEWEB)

    Burgard, K.C.

    1998-04-09

    The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis.

  10. Evaluation of health and safety impacts of defense high-level waste in geologic repositories

    International Nuclear Information System (INIS)

    Smith, E.D.; Kocher, D.C.; Witherspoon, J.P.

    1985-02-01

    Pursuant to the requirement of the Nuclear Waste Policy Act of 1982 that the President evaluate the use of commercial high-level waste repositories for the disposal of defense high-level wastes, a comparative assessment has been performed of the potential health and safety impacts of disposal of defense wastes in commercial or defense-only repositories. Simplified models were used to make quantitative estimates of both long- and short-term health and safety impacts of several options for defense high-level waste disposal. The results indicate that potential health and safety impacts are not likely to vary significantly among the different disposal options for defense wastes. Estimated long-term health and safety impacts from all defense-waste disposal options are somewhat less than those from commercial waste disposal, while short-term health and safety impacts appear to be insensitive to the differences between defense and commercial wastes. In all cases, potential health and safety impacts are small because of the need to meet stringent standards promulgated by the US Environmental Protection Agency and the US Nuclear Regulatory Commission. We conclude that health and safety impacts should not be a significant factor in the choice of a disposal option for defense high-level wastes. 20 references, 14 tables

  11. Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)

    Energy Technology Data Exchange (ETDEWEB)

    Burgard, K.C.

    1998-06-02

    The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis.

  12. Radionuclide compositions of spent fuel and high level waste from commercial nuclear reactors

    International Nuclear Information System (INIS)

    Goodill, D.R.; Tymons, B.J.

    1984-10-01

    This report provides information on radionuclide compositions of spent fuel and high level waste produced during reprocessing. The reactor types considered are Magnox, AGR, PWR and CFR. The activities of the radionuclides are calculated using the FISPIN code. The results are presented in a form suitable for radioactive waste management calculations. (author)

  13. Safety of handling, storing and transportation of spent nuclear fuel and vitrified high-level wastes

    International Nuclear Information System (INIS)

    Ericsson, A.M.

    1977-11-01

    The safety of handling and transportation of spent fuel and vitrified high-level waste has been studied. Only the operations which are performed in Sweden are included. That is: - Transportation of spent fuel from the reactors to an independant spent fuel storage installation (ISFSI). - Temporary storage of spent fuel in the ISFSI. - Transportation of the spent fuel from the ISFSI to a foreign reprocessing plant. - Transportation of vitrified high-level waste to an interim storage facility. - Interim storage of vitrified high-level waste. - Handling of the vitrified high-level waste in a repository for ultimate disposal. For each stage in the handling sequence above the following items are given: - A brief technical description. - A description of precautionary measures considered in the design. - An analysis of the discharges of radioactive materials to the environment in normal operation. - An analysis of the discharges of radioactive materials due to postulated accidents. The dose to the public has been roughly and conservatively estimated for both normal and accident conditions. The expected rate of occurence are given for the accidents. The results show that above described handling sequence gives only a minor risk contribution to the public

  14. Tank safety screening data quality objective. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, J.W.

    1995-04-27

    The Tank Safety Screening Data Quality Objective (DQO) will be used to classify 149 single shell tanks and 28 double shell tanks containing high-level radioactive waste into safety categories for safety issues dealing with the presence of ferrocyanide, organics, flammable gases, and criticality. Decision rules used to classify a tank as ``safe`` or ``not safe`` are presented. Primary and secondary decision variables used for safety status classification are discussed. The number and type of samples required are presented. A tabular identification of each analyte to be measured to support the safety classification, the analytical method to be used, the type of sample, the decision threshold for each analyte that would, if violated, place the tank on the safety issue watch list, and the assumed (desired) analytical uncertainty are provided. This is a living document that should be evaluated for updates on a semiannual basis. Evaluation areas consist of: identification of tanks that have been added or deleted from the specific safety issue watch lists, changes in primary and secondary decision variables, changes in decision rules used for the safety status classification, and changes in analytical requirements. This document directly supports all safety issue specific DQOs and additional characterization DQO efforts associated with pretreatment and retrieval. Additionally, information obtained during implementation can assist in resolving assumptions for revised safety strategies, and in addition, obtaining information which will support the determination of error tolerances, confidence levels, and optimization schemes for later revised safety strategy documentation.

  15. Removal of actinides from high-level wastes generated in the reprocessing of commercial fuels

    International Nuclear Information System (INIS)

    Bond, W.D.; Leuze, R.E.

    1975-09-01

    Progress is reported on a technical feasibility study of removing the very long-lived actinides (uranium, neptunium, plutonium, americium, and curium) from high-level wastes generated in the commercial reprocessing of spent nuclear fuels. The study was directed primarily at wastes from the reprocessing of light water reactor (LWR) fuels and specifically to developing satisfactory methods for reducing the actinide content of these wastes to values that would make 1000-year-decayed waste comparable in radiological toxicity to natural uranium ore deposits. Although studies are not complete, results thus far indicate the most promising concept for actinide removal includes both improved recovery of actinides in conventional fuel reprocessing and secondary processing of the high-level wastes. Secondary processing will be necessary for the removal of americium and curium and perhaps some residual plutonium. Laboratory-scale studies of separations methods that appear most promising are reported and conceptual flowsheets are discussed. (U.S.)

  16. Transferring knowledge about high-level waste repositories: An ethical consideration

    International Nuclear Information System (INIS)

    Berndes, S.; Kornwachs, K.

    1996-01-01

    The purpose of this paper is to present requirements to Information and Documentation Systems for high-level waste repositories from an ethical point of view. A structured synopsis of ethical arguments used by experts from Europe and America is presented. On the one hand the review suggests to reinforce the obligation to transfer knowledge about high level waste repositories. This obligation is reduced on the other hand by the objection that ethical obligations are dependent on the difference between our and future civilizations. This reflection results in proposing a list of well-balanced ethical arguments. Then a method is presented which shows how scenarios of possible future civilizations for different time horizons and related ethical arguments are used to justify requirements to the Information and Documentation System

  17. Remote Handling Equipment for a High-Level Waste Package Closure System

    International Nuclear Information System (INIS)

    Kevin M. Croft; Scott M. Allen; Mark W. Borland

    2006-01-01

    High-level waste will be placed in sealed waste packages inside a shielded closure cell. The Idaho National Laboratory (INL) has designed a system for closing the waste packages including all cell interior equipment and support systems. This paper discusses the material handling aspects of the equipment used and operations that will take place as part of the waste package closure operations. Prior to construction, the cell and support system will be assembled in a full-scale mockup at INL

  18. Mercury Phase II Study - Mercury Behavior across the High-Level Waste Evaporator System

    Energy Technology Data Exchange (ETDEWEB)

    Bannochie, C. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Crawford, C. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jackson, D. G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Shah, H. B. [Savannah River Remediation, LLC., Aiken, SC (United States); Jain, V. [Savannah River Remediation, LLC., Aiken, SC (United States); Occhipinti, J. E. [Savannah River Remediation, LLC., Aiken, SC (United States); Wilmarth, W. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-06-17

    The Mercury Program team’s effort continues to develop more fundamental information concerning mercury behavior across the liquid waste facilities and unit operations. Previously, the team examined the mercury chemistry across salt processing, including the Actinide Removal Process/Modular Caustic Side Solvent Extraction Unit (ARP/MCU), and the Defense Waste Processing Facility (DWPF) flowsheets. This report documents the data and understanding of mercury across the high level waste 2H and 3H evaporator systems.

  19. Geological aspects of the high level waste and spent fuel disposal programme in Slovakia

    Energy Technology Data Exchange (ETDEWEB)

    Matej, Gedeon; Milos, Kovacik; Jozef, Hok [Geological Survey of Slovak Republic, Bratislava (Slovakia)

    2001-07-01

    An autonomous programme for development of a deep geological high level waste and spent fuel disposal began in 1996. One of the most important parts in the programme is siting of the future deep seated disposal. Geological conditions in Slovakia are complex due to the Alpine type tectonics that formed the geological environment during Tertiary. Prospective areas include both crystalline complexes (tonalites, granites, granodiorites) and Neogene (Miocene) argillaceous complexes. (author)

  20. Numerical aspects of the modelling of the local effects of a high level waste repository

    International Nuclear Information System (INIS)

    Ferreri, J.C.; Ventura, M.A.

    1985-01-01

    The numerical approximations adapted for the development of the computational models for the prediction of the effects of the emplacement of a high level waste repository are reviewed. The problems considered include: the thermal history of the rocky mass constituting the burial media, the flow of underground water and the associated migration of radionuclides in the same media. Results associated with verification of the implemented codes are presented. Their limitations and advantages are discussed. (Author) [es

  1. Separation of transuranium elements from high-level waste by extraction with diisodecyl phosphoric acid

    International Nuclear Information System (INIS)

    Morita, Y.; Kubota, M.; Tani, S.

    1991-01-01

    Separation of transuranic elements (TRU) by extraction with diisodecyl phosphoric acid (DIDPA) has been studied to develop a partitioning process for high-level waste (HLW). In the present study, experiments of counter-current continuous extraction and back-extraction using a miniature mixer-settler were carried out to find the optimum process condition for the separation of Np initially in the pentavalent state and to examine the extraction behaviors of fission and corrosion products. (J.P.N.)

  2. Remote Handling Equipment for a High-Level Waste Waste Package Closure System

    Energy Technology Data Exchange (ETDEWEB)

    Kevin M. Croft; Scott M. Allen; Mark W. Borland

    2006-04-01

    High-level waste will be placed in sealed waste packages inside a shielded closure cell. The Idaho National Laboratory (INL) has designed a system for closing the waste packages including all cell interior equipment and support systems. This paper discusses the material handling aspects of the equipment used and operations that will take place as part of the waste package closure operations. Prior to construction, the cell and support system will be assembled in a full-scale mockup at INL.

  3. Development of site suitability criteria for the high level waste repository for Lawrence Livermore Laboratories

    International Nuclear Information System (INIS)

    1977-06-01

    Results of our mining, geological and geotechnical studies provided in support of the development of site suitability criteria for the high level waste repository are presented. The primary purpose of the work was the identification and development of appropriate geotechnical descriptors and coefficients required for the Site Suitability Repository Model. This model was developed by The Analytic Sciences Corporation (TASC) of Reading, Massachusetts and is not described in this report

  4. Separation of palladium from high-level waste using metal ferro cyanide loaded resins

    International Nuclear Information System (INIS)

    Valsala, T.P.; Joseph, Annie; Yeotikar, R.G.

    2005-01-01

    High-level waste (HLW) is generated during reprocessing of spent fuel. HLW contains corrosion products, unextracted actinides, process chemicals and fission products. A recent trend is there to consider waste as a source of wealth. Among the fission products separation and recovery of platinum group metals have gained great attention. HLW is a good source of palladium of the platinum group metal. The present study shows the feasibility of ion exchange separation of Pd from HLW. (author)

  5. Effects of heat from high-level waste on performance of deep geological repository components

    International Nuclear Information System (INIS)

    1984-11-01

    This report discusses the effects of heat on the deep geological repository systems and its different components. The report is focussed specifically on effects due to thermal energy release solely from high-level waste or spent fuel. It reviews the experimental data and theoretical models of the effects of heat both on the behaviour of engineered and natural barriers. A summary of the current status of research and repository development including underground test facilities is presented

  6. Cost calculation and financial measures for high-level waste disposal business

    International Nuclear Information System (INIS)

    Sekiguchi, Hiromasa.

    1987-01-01

    A study is made on the costs for disposal of high-level wastes, centering on financial problems involving cost calculation for disposal business and methods and systems for funding the business. The first half of the report is focused on calculation of costs for disposal business. Basic equations are shown to calculate the total costs required for a disposal plant and the costs for disposal of one unit of high-level wastes. A model is proposed to calculate the charges to be paid by electric power companies to the plant for disposal of their wastes. Another equation is derived to calculate the disposal charge per kWh of power generation in a power plant. The second half of the report is focused on financial measures concerning expenses for disposal. A financial basis should be established for the implementation of high-level waste disposal. It is insisted that a reasonable method for estimating the disposal costs should be set up and it should be decided who will pay the expenses. Discussions are made on some methods and systems for funding the disposal business. An additional charge should be included in the electricity bill to be paid by electric power users, or it should be included in tax. (Nogami, K.)

  7. DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

    International Nuclear Information System (INIS)

    Radulesscu, G.; Tang, J.S.

    2000-01-01

    The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M andO [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M andO 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M andQ 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M andO 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable canisters. The intended use of this

  8. DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

    Energy Technology Data Exchange (ETDEWEB)

    G. Radulesscu; J.S. Tang

    2000-06-07

    The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M&O 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M&Q 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M&O 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable

  9. Crystal accumulation in the Hanford Waste Treatment Plant high level waste melter. Preliminary settling and resuspension testing

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Fowley, M. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Miller, D. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-05-01

    The full-scale, room-temperature Hanford Tank Waste Treatment and Immobilization Plant (WTP) High-Level Waste (HLW) melter riser test system was successfully operated with silicone oil and magnetite particles at a loading of 0.1 vol %. Design and construction of the system and instrumentation, and the selection and preparation of simulant materials, are briefly reviewed. Three experiments were completed. A prototypic pour rate was maintained, based on the volumetric flow rate. Settling and accumulation of magnetite particles were observed at the bottom of the riser and along the bottom of the throat after each experiment. The height of the accumulated layer at the bottom of the riser, after the first pouring experiment, approximated the expected level given the solids loading of 0.1 vol %. More detailed observations of particle resuspension and settling were made during and after the third pouring experiment. The accumulated layer of particles at the bottom of the riser appeared to be unaffected after a pouring cycle of approximately 15 minutes at the prototypic flow rate. The accumulated layer of particles along the bottom of the throat was somewhat reduced after the same pouring cycle. Review of the time-lapse recording showed that some of the settling particles flow from the riser into the throat. This may result in a thicker than expected settled layer in the throat.

  10. Microstructural properties of high level waste concentrates and gels with raman and infrared spectroscopies. 1997 annual progress report

    International Nuclear Information System (INIS)

    Agnew, S.F.; Coarbin, R.A.; Johnston, C.T.

    1997-01-01

    'Monosodium aluminate, the phase of aluminate found in waste tanks, is only stable over a fairly narrow range of water vapor pressure (22% relative humidity at 22 C). As a result, aluminate solids are stable at Hanford (seasonal average RH ∼20%) but are not be stable at Savannah River (seasonal average RH ∼40%). Monosodium aluminate (MSA) releases water upon precipitation from solution. In contrast, trisodium aluminate (TSA) consumes water upon precipitation. As a result, MSA precipitates gradually over time while TSA undergoes rapid accelerated precipitation, often gelling its solution. Raman spectra reported for first time for monosodium and trisodium aluminate solids. Ternary phase diagrams can be useful for showing effects of water removal, even with concentrated waste. Kinetics of monosodium aluminate precipitation are extremely slow (several months) at room temperature but quite fast (several hours) at 60 C. As a result, all waste simulants that contain aluminate need several days of cooking at 60 C in order to truly represent the equilibrium state of aluminate. The high level waste (HLW) slurries that have been created at the Hanford and Savannah River Sites over that last fifty years constitute a large fraction of the remaining HLW volumes at both sites. In spite of the preponderance of these wastes, very little quantitative information is available about their physical and chemical properties other than elemental analyses.'

  11. Design and operating features of the high-level waste vitrification system for the West Valley demonstration project

    Energy Technology Data Exchange (ETDEWEB)

    Siemens, D.H.; Beary, M.M.; Barnes, S.M.; Berger, D.N.; Brouns, R.A.; Chapman, C.C.; Jones, R.M.; Peters, R.D.; Peterson, M.E.

    1986-03-01

    A liquid-fed joule-heated ceramic melter system is the reference process for immobilization of the high-level liquid waste in the US and several foreign countries. This system has been under development for over ten years at Pacific Northwest Laboratory and other national laboratories operated for the US Department of Energy. Pacific Northwest Laboratory contributed to this research through its Nuclear Waste Treatment Program and used applicable data to design and test melters and related systems using remote handling of simulated radioactive wastes. This report describes the equipment designed in support of the high-level waste vitrification program at West Valley, New York. Pacific Northwest Laboratory worked closely with West Valley Nuclear Services Company to design a liquid-fed ceramic melter, a liquid waste preparation and feed tank and pump, an off-gas treatment scrubber, and an enclosed turntable for positioning the waste canisters. Details of these designs are presented including the rationale for the design features and the alternatives considered.

  12. Process description and plant design for preparing ceramic high-level waste forms

    International Nuclear Information System (INIS)

    Grantham, L.F.; McKisson, R.L.; Guon, J.; Flintoff, J.F.; McKenzie, D.E.

    1983-01-01

    The ceramics process flow diagram has been simplified and upgraded to utilize only two major processing steps - fluid-bed calcination and hot isostatic press consolidating. Full-scale fluid-bed calcination has been used at INEL to calcine high-level waste for 18 y; and a second-generation calciner, a fully remotely operated and maintained calciner that meets ALARA guidelines, started calcining high-level waste in 1982. Full-scale hot isostatic consolidation has been used by DOE and commercial enterprises to consolidate radioactive components and to encapsulate spent fuel elements for several years. With further development aimed at process integration and parametric optimization, the operating knowledge of full-scale demonstration of the key process steps should be rapidly adaptable to scale-up of the ceramic process to full plant size. Process flowsheets used to prepare ceramic and glass waste forms from defense and commercial high-level liquid waste are described. Preliminary layouts of process flow diagrams in a high-level processing canyon were prepared and used to estimate the preliminary cost of the plant to fabricate both waste forms. The estimated costs for using both options were compared for total waste management costs of SRP high-level liquid waste. Using our design, for both the ceramic and glass plant, capital and operating costs are essentially the same for both defense and commercial wastes, but total waste management costs are calculated to be significantly less for defense wastes using the ceramic option. It is concluded from this and other studies that the ceramic form may offer important advantages over glass in leach resistance, waste loading, density, and process flexibility. Preliminary economic calculations indicate that ceramics must be considered a leading candidate for the form to immobilize high-level wastes

  13. Partitioning and recovery of neptunium from high level waste streams of PUREX origin using 30% TBP

    International Nuclear Information System (INIS)

    Mathur, J.N.; Murali, M.S.; Balarama Krishna, M.V.; Iyer, R.H.; Chitnis, R.R.; Wattal, P.K.; Theyyunni, T.K.; Ramanujam, A.; Dhami, P.S.; Gopalakrishnan, V.

    1995-01-01

    237 Np is one of the longest-lived nuclides among the actinides present in the high level waste solutions of reprocessing origin. Its separation, recovery and transmutation can reduce the problem of long term storage of the vitrified waste to a great extent. With this objective, the present work was initiated to study the extraction of neptunium into TBP under the conditions relevant to high level waste, along with uranium and plutonium by oxidising it to hexavalent state using potassium dichromate and subsequently recovering it by selective stripping. Three types of simulated HLW solutions namely sulphate bearing (SB), with an acidity of ∼ 0.3 M and non-sulphate wastes originating from the reprocessing of fuels from pressurised heavy water reactor (PHWR) and fast breeder reactor (FBR) with acidities of 3.0 M HNO 3 were employed in these studies. The extraction of U(VI), Np(VI) and Pu(VI) was very high for PHWR- and FBR-HLW solutions, whereas for the SB-HLW solution, these values were less but reasonably high. Quantitative recovery of neptunium and plutonium was achieved using a stripping solution containing 0.1 M H 2 O 2 and 0.01 M ascorbic acid at an acidity of 2.0 M. Since, cerium present in the waste solutions is expected to undergo oxidation in presence of K 2 Cr 2 O 7 , its extraction behaviour was also studied under similar conditions. Based on the results, a scheme was formulated for the recovery of neptunium along with plutonium and was successfully applied to actual high level waste solution originating from the reprocessing of research reactor fuels. (author). 19 refs., 2 figs., 17 tabs

  14. An analytical hierarchy process for decision making of high-level-waste management

    International Nuclear Information System (INIS)

    Wang, J.H.C.; Jang, W.

    1995-01-01

    To prove the existence value of nuclear technology for the world of post cold war, demonstration of safe rad-waste disposal is essential. High-level-waste (HLW) certainly is the key issue to be resolved. To assist a rational and persuasive process on various disposal options, an Analytical Hierarchy Process (AHP) for the decision making of HLW management is presented. The basic theory and rationale are discussed, and applications are shown to illustrate the usefulness of the AHP. The authors wish that the AHP can provide a better direction for the current doomed situations of Taiwan nuclear industry, and to exchange with other countries for sharing experiences on the HLW management

  15. CREVICE CORROSION and PITTING OF HIGH-LEVEL WASTE CONTAINERS: INTEGRATION OF DETERMINISTIC and PROBABILISTIC MODELS

    International Nuclear Information System (INIS)

    JOSEPH C. FARMER AND R. DANIEL MCCRIGHT

    1997-01-01

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological repository at Yucca Mountain, Nevada is a two-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as carbon steel or Monel 400. An integrated predictive model is being developed to account for the effects of localized environmental conditions in the CRM-CAM crevice on the initiation and propagation of pits through the CRM

  16. Evaluation of high-level waste pretreatment processes with an approximate reasoning model

    International Nuclear Information System (INIS)

    Bott, T.F.; Eisenhawer, S.W.; Agnew, S.F.

    1999-01-01

    The development of an approximate-reasoning (AR)-based model to analyze pretreatment options for high-level waste is presented. AR methods are used to emulate the processes used by experts in arriving at a judgment. In this paper, the authors first consider two specific issues in applying AR to the analysis of pretreatment options. They examine how to combine quantitative and qualitative evidence to infer the acceptability of a process result using the example of cesium content in low-level waste. They then demonstrate the use of simple physical models to structure expert elicitation and to produce inferences consistent with a problem involving waste particle size effects

  17. Evaluation of high-level waste pretreatment processes with an approximate reasoning model

    Energy Technology Data Exchange (ETDEWEB)

    Bott, T.F.; Eisenhawer, S.W.; Agnew, S.F.

    1999-04-01

    The development of an approximate-reasoning (AR)-based model to analyze pretreatment options for high-level waste is presented. AR methods are used to emulate the processes used by experts in arriving at a judgment. In this paper, the authors first consider two specific issues in applying AR to the analysis of pretreatment options. They examine how to combine quantitative and qualitative evidence to infer the acceptability of a process result using the example of cesium content in low-level waste. They then demonstrate the use of simple physical models to structure expert elicitation and to produce inferences consistent with a problem involving waste particle size effects.

  18. High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 3

    International Nuclear Information System (INIS)

    Cunnane, J.C.

    1994-03-01

    The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II

  19. High Level Waste plant operation and maintenance concepts. Final report, March 27, 1995

    International Nuclear Information System (INIS)

    Janicek, G.P.

    1995-01-01

    The study reviews and evaluates worldwide High Level Waste (HLW) vitrification operating and maintenance (O ampersand M) philosophies, plant design concepts, and lessons learned with an aim towards developing O ampersand M recommendations for either, similar implementation or further consideration in a HLW vitrification facility at Hanford. The study includes a qualitative assessment of alternative concepts for a variety of plant and process systems and subsystems germane to HLW vitrification, such as, feed materials handling, melter configuration, glass form, canister handling, failed equipment handling, waste handling, and process control. Concept evaluations and recommendations consider impacts to Capital Cost, O ampersand M Cost, ALARA, Availability, and Reliability

  20. CLASSIFICATION OF THE MGR DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER SYSTEM

    International Nuclear Information System (INIS)

    J.A. Ziegler

    1999-01-01

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) defense high-level waste disposal container system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333PY ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998)

  1. Validation of the Performance of High-level Waste Disposal System

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Won Jin; Park, J. H.; Lee, J. O. (and others)

    2007-06-15

    The experimental researches to validate the integrity and safety of high-level waste disposal system were carried out. The studies on the construction of KURT, and the site rock characteristics were conducted. Thermal-hydro-mechanical behavior of engineered barrier system was investigated using the engineering-scale test facility. The migration and retardation of radionuclide through the rock fracture under anaerobic and reducing condition were studied. The distribution coefficients of radionuclides onto granite, the rock matrix diffusion coefficients, and the gap and grain boundary inventories of spent fuel were measured.

  2. High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    Cunnane, J.C. [comp.; Bates, J.K.; Bradley, C.R. [Argonne National Lab., IL (United States)] [and others

    1994-03-01

    The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II.

  3. Fracture toughness measurements on a glass bonded sodalite high-level waste form

    International Nuclear Information System (INIS)

    DiSanto, T.; Goff, K. M.; Johnson, S. G.; O'Holleran, T. P.

    1999-01-01

    The electrometallurgical treatment of metallic spent nuclear fuel produces two high-level waste streams; cladding hulls and chloride salt. Argonne National Laboratory is developing a glass bonded sodalite waste form to immobilize the salt waste stream. The waste form consists of 75 Vol.% crystalline sodalite (containing the salt) with 25 Vol.% of an ''intergranular'' glassy phase. Microindentation fracture toughness measurements were performed on representative samples of this material using a Vickers indenter. Palmqvist cracking was confirmed by post-indentation polishing of a test sample. Young's modulus was measured by an acoustic technique. Fracture toughness, microhardness, and Young's modulus values are reported, along with results from scanning electron microscopy studies

  4. Preliminary evaluation of alternative forms for immobilization of Savannah River Plant high-level waste

    International Nuclear Information System (INIS)

    Stone, J.A.; Goforth, S.T. Jr.; Smith, P.K.

    1979-12-01

    An evaluation of available information on eleven alternative solid forms for immobilization of SRP high-level waste has been completed. Based on the assessment of both product and process characteristics, four forms were selected for more detailed evaluation: (1) borosilicate glass made in the reference process, (2) a high-silica glass made from a porous glass matrix, (3) crystalline ceramics such as supercalcine or SYNROC, and (4) ceramics coated with an impervious barrier. The assessment includes a discussion of product and process characteristics for each of the eleven forms, a cross comparison of these characteristics for the forms, and the bases for selecting the most promising forms for further study

  5. The feasibility of sampling the glass pour in a high level waste vitrification plant

    International Nuclear Information System (INIS)

    Cole, G.V.; Shilton, P.; Morris, J.B.

    1986-06-01

    Vitrified high level waste can be sampled for quality assurance purposes in three general ways: (I) from the glass pour, (II) from the canister, and (III) from the melter. A discussion of the potential advantages and disadvantages of each route is presented. The second philosophy seems to show the best promise; it is recommended that the Contained Pot method and the Token method are best suited for further development. An international survey of policy at vitrification plants shows that with one possible exception no glass sampling is intended and that quality is normally to be assured by control of the vitrification process. (author)

  6. Format and content for the license application for the high-level waste repository

    International Nuclear Information System (INIS)

    1990-11-01

    The purpose of this regulatory guide, ''Format and Content for the License Application for the High-Level Waste Repository,'' is to indicate the information to be provided in the LA and to establish a format acceptable to the NRC staff for presenting the information. Use of this format will help ensure the completeness of the information provided, will assist the NRC staff and others in locating the information, and will shorten the time needed for the review process. This guide is being issued as a draft to afford DOE, government agencies, and members of the public an opportunity to submit comments on the guide

  7. Crystal accumulation in the Hanford Waste Treatment Plant high level waste melter: Summary of FY2016 experiements

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. [Savannah River Site (SRS), Aiken, SC (United States); Fowley, M. [Savannah River Site (SRS), Aiken, SC (United States); Miller, D. [Savannah River Site (SRS), Aiken, SC (United States)

    2016-12-01

    Five experiments were completed with the full-scale, room temperature Hanford Waste Treatment and Immobilization Plant (WTP) high-level waste (HLW) melter riser test system to observe particle flow and settling in support of a crystal tolerant approach to melter operation. A prototypic pour rate was maintained based on the volumetric flow rate. Accumulation of particles was observed at the bottom of the riser and along the bottom of the throat after each experiment. Measurements of the accumulated layer thicknesses showed that the settled particles at the bottom of the riser did not vary in thickness during pouring cycles or idle periods. Some of the settled particles at the bottom of the throat were re-suspended during subsequent pouring cycles, and settled back to approximately the same thickness after each idle period. The cause of the consistency of the accumulated layer thicknesses is not year clear, but was hypothesized to be related to particle flow back to the feed tank. Additional experiments reinforced the observation of particle flow along a considerable portion of the throat during idle periods. Limitations of the system are noted in this report and may be addressed via future modifications. Follow-on experiments will be designed to evaluate the impact of pouring rate on particle re-suspension, the influence of feed tank agitation on particle accumulation, and the effect of changes in air lance positioning on the accumulation and re-suspension of particles at the bottom of the riser. A method for sampling the accumulated particles will be developed to support particle size distribution analyses. Thicker accumulated layers will be intentionally formed via direct addition of particles to select areas of the system to better understand the ability to continue pouring and re-suspend particles. Results from the room temperature system will be correlated with observations and data from the Research Scale Melter (RSM) at Pacific Northwest National Laboratory

  8. Assessment of the radiological protection aspects of disposal of high level waste on the ocean floor

    International Nuclear Information System (INIS)

    Grimwood, P.D.; Webb, G.A.M.

    1976-10-01

    This study is a preliminary assessment of the potential radiological consequences of disposal of solidified high-level radioactive waste on the floor of the deep ocean. As an input to the modelling used in the assessment, an arbitrary choice is made to consider the total high-level waste which would be generated by a postulated world nuclear power programme to the year 2000. It is assumed that all this waste, in solidified form, is disposed of on to the floor of the North Atlantic. The body of this report is the modelling of the subsequent release of activity into the water, its dispersion in the ocean and eventual uptake in marine organisms and sediments. The consequent radiation exposure of man is assessed in terms of both individual and collective doses. It is intended that only broad conclusions should be drawn from this study. The objective of the assessment is to highlight those subject areas where more study of information is required before a decision can be reached regarding this method of disposal. No overriding reason connected with the radiological protection considerations has been identified which would preclude the disposal of suitably conditioned high-level waste on the ocean floor. Further evaluation of this disposal option is therefore justified. (author)

  9. Long-term high-level waste technology. Composite quarterly technical report: April-June 1981

    International Nuclear Information System (INIS)

    Cornman, W.R.

    1981-12-01

    This series of reports summarizes research and development studies on the immobilization of high-level wastes from the chemical reprocessing of nuclear reactor fuels. The reports are grouped under the following tasks: (1) program management and support; (2) waste preparation; (3) waste fixation; and (4) final handling. Some of the highlights are: leaching properties were obtained for titanate and tailored ceramic materials being developed at ICPP to immobilize zirconia calcine; comparative leach tests, hot-cell tests, and process evaluations were conducted of waste form alternatives to borosilicate glass for the immobilization of SRP high-level wastes, experiments were run at ANL to qualify neutron activation analysis and radioactive tracers for measuring leach rates from simulated waste glasses; comparative leach test samples of SYNROC D were prepared, characterized, and tested at LLNL; encapsulation of glass marbles with lead or lead alloys was demonstrated on an engineering scale at PNL; a canister for reference Commercial HLW was designed at PNL; a study of the optimization of salt-crete was completed at SRL; a risk assessment showed that an investment for tornado dampers in the interim storage building of the DWPF is unjustified

  10. High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 2

    International Nuclear Information System (INIS)

    Cunnane, J.C.

    1994-03-01

    The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion.This document is organized into three volumes. Volumes I and II represent a tiered set of information intended for somewhat different audiences. Volume I is intended to provide an overview of waste glass corrosion, and Volume 11 is intended to provide additional experimental details on experimental factors that influence waste glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II. Volume I is intended for managers, decision makers, and modelers, the combined set of Volumes I, II, and III is intended for scientists and engineers working in the field of high-level waste

  11. High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Cunnane, J.C. [comp.; Bates, J.K.; Bradley, C.R. [Argonne National Lab., IL (United States)] [and others

    1994-03-01

    The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion.This document is organized into three volumes. Volumes I and II represent a tiered set of information intended for somewhat different audiences. Volume I is intended to provide an overview of waste glass corrosion, and Volume 11 is intended to provide additional experimental details on experimental factors that influence waste glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II. Volume I is intended for managers, decision makers, and modelers, the combined set of Volumes I, II, and III is intended for scientists and engineers working in the field of high-level waste.

  12. Quality assurance requirements and methods for high level waste package acceptability

    International Nuclear Information System (INIS)

    1992-12-01

    This document should serve as guidance for assigning the necessary items to control the conditioning process in such a way that waste packages are produced in compliance with the waste acceptance requirements. It is also provided to promote the exchange of information on quality assurance requirements and on the application of quality assurance methods associated with the production of high level waste packages, to ensure that these waste packages comply with the requirements for transportation, interim storage and waste disposal in deep geological formations. The document is intended to assist both the operators of conditioning facilities and repositories as well as national authorities and regulatory bodies, involved in the licensing of the conditioning of high level radioactive wastes or in the development of deep underground disposal systems. The document recommends the quality assurance requirements and methods which are necessary to generate data for these parameters identified in IAEA-TECDOC-560 on qualitative acceptance criteria, and indicates where and when the control methods can be applied, e.g. in the operation or commissioning of a process or in the development of a waste package design. Emphasis is on the control of the process and little reliance is placed on non-destructive or destructive testing. Qualitative criteria, relevant to disposal of high level waste, are repository dependent and are not addressed here. 37 refs, 3 figs, 2 tabs

  13. Preparation of plutonium waste forms with ICPP calcined high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Staples, B.A.; Knecht, D.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States); O`Holleran, T.P. [Argonne National Lab.-West, Idaho Falls, ID (United States)] [and others

    1997-05-01

    Glass and glass-ceramic forms developed for the immobilization of calcined high-level wastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce{sup +4}) as a surrogate for plutonium (Pu{sup +4}) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-level waste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study.

  14. Techno-economical Analysis of High Level Waste Storage and Disposal Options

    International Nuclear Information System (INIS)

    Bace, M.; Trontl, K.; Vrankic, K.

    2002-01-01

    Global warming and instability of gas and oil prices are redefining the role of nuclear energy in electrical energy production. A production of high-level radioactive waste (HLW), during the nuclear power plant operation and a danger of high level waste mitigation to the environment are considered by the public as a main obstacle of accepting the nuclear option. As economical and technical aspects of the back end of fuel cycle will affect the nuclear energy acceptance the techno-economical analysis of different methods for high level waste storage and disposal has to be performed. The aim of this paper is to present technical and economical characteristics of different HLW storage and disposal technologies. The final choice of a particular HLW management method is closely connected to the selection of a fuel cycle type: open or closed. Wet and dry temporary storage has been analyzed including different types of spent fuel pool capacity increase methods, different pool location (at reactor site and away from reactor site) as well as casks and vault system of dry storage. Since deep geological deposition is the only disposal method with a realistic potential, we focused our attention on that disposal technology. Special attention has been given to the new idea of international and regional disposal location. The analysis showed that a coexistence of different storage methods and deep geological deposition is expected in the future, regardless of the fuel cycle type. (author)

  15. International safeguards relevant to geologic disposal of high-level wastes and spent fuels

    International Nuclear Information System (INIS)

    Pillay, K.K.S.; Picard, R.R.

    1989-01-01

    Spent fuels from once-through fuel cycles placed in underground repositories have the potential to become attractive targets for diversion and/or theft because of their valuable material content and decreasing radioactivity. The first geologic repository in the US, as currently designed, will contain approximately 500 Mt of plutonium, 60,000 Mt of uranium and a host of other fissile and strategically important elements. This paper identifies some of the international safeguards issues relevant to the various proposed scenarios for disposing of the spent fuel. In the context of the US program for geologic disposal of spent fuels, this paper highlights several issues that should be addressed in the near term by US industries, the Department of Energy, and the Nuclear Regulatory Commission before the geologic repositories for spent fuels become a reality. Based on US spent fuel discharges, an example is presented to illustrate the enormity of the problem of verifying spent fuel inventories. The geologic disposal scenario for high-level wastes originating from defense facilities produced a ''practicably irrecoverable'' waste form. Therefore, safeguards issues for geologic disposal of high-level waste now in the US are less pressing. 56 refs. , 2 figs

  16. Preparation of plutonium waste forms with ICPP calcined high-level waste

    International Nuclear Information System (INIS)

    Staples, B.A.; Knecht, D.A.; O'Holleran, T.P.

    1997-05-01

    Glass and glass-ceramic forms developed for the immobilization of calcined high-level wastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce +4 ) as a surrogate for plutonium (Pu +4 ) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-level waste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study

  17. Crystallization in high-level waste glass: A review of glass theory and noteworthy literature

    Energy Technology Data Exchange (ETDEWEB)

    Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-01

    There is a fundamental need to continue research aimed at understanding nepheline and spinel crystal formation in high-level waste (HLW) glass. Specifically, the formation of nepheline solids (K/NaAlSiO₄) during slow cooling of HLW glass can reduce the chemical durability of the glass, which can cause a decrease in the overall durability of the glass waste form. The accumulation of spinel solids ((Fe, Ni, Mn, Zn)(Fe,Cr)₂O₄), while not detrimental to glass durability, can cause an array of processing problems inside of HLW glass melters. In this review, the fundamental differences between glass and solid-crystals are explained using kinetic, thermodynamic, and viscosity arguments, and several highlights of glass-crystallization research, as it pertains to high-level waste vitrification, are described. In terms of mitigating spinel in the melter and both spinel and nepheline formation in the canister, the complexity of HLW glass and the intricate interplay between thermal, chemical, and kinetic factors further complicates this understanding. However, new experiments seeking to elucidate the contributing factors of crystal nucleation and growth in waste glass, and the compilation of data from older experiments, may go a long way towards helping to achieve higher waste loadings while developing more efficient processing strategies.

  18. Immobilization of defense high-level waste: an assessment of technological strategies and potential regulatory goals. Volume II

    International Nuclear Information System (INIS)

    1979-06-01

    This volume contains the following appendices: selected immobilization processes, directory of selected European organizations involved in HLW management, U.S. high-level waste inventories, and selected European HLW program

  19. MAR Assessments Of The High Level Waste System Plan Revision 16

    International Nuclear Information System (INIS)

    Peeler, D.; Edwards, T.

    2011-01-01

    High-level waste (HLW) throughput (i.e., the amount of waste processed per unit of time) is primarily a function of two critical parameters: waste loading (WL) and melt rate. For the Defense Waste Processing Facility (DWPF), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). Significant increases in waste throughput have been achieved at DWPF since initial radioactive operations began in 1996. Key technical and operational initiatives that supported increased waste throughput included improvements in facility attainment, the Chemical Processing Cell (CPC) flowsheet, process control models and frit formulations. As a result of these key initiatives, DWPF increased WLs from a nominal 28% for Sludge Batch 2 (SB2) to ∼34 to 38% for SB3 through SB6 while maintaining or slightly improving canister fill times. Although considerable improvements in waste throughput have been obtained, future contractual waste loading targets are nominally 40%, while canister production rates are also expected to increase (to a rate of 325 to 400 canisters per year). Although implementation of bubblers have made a positive impact on increasing melt rate for recent sludge batches targeting WLs in the mid30s, higher WLs will ultimately make the feeds to DWPF more challenging to process. Savannah River Remediation (SRR) recently requested the Savannah River National Laboratory (SRNL) to perform a paper study assessment using future sludge projections to evaluate whether the current Process Composition Control System (PCCS) algorithms would provide projected operating windows to allow future contractual WL targets to be met. More specifically, the objective of this study was to evaluate future sludge batch projections (based on Revision 16 of the HLW Systems Plan) with respect to projected operating windows using current PCCS models and associated constraints. Based on the assessments, the waste loading interval over

  20. Crystallization In High Level Waste (HLW) Glass Melters: Operational Experience From The Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-02-27

    processing strategy for the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The basis of this alternative approach is an empirical model predicting the crystal accumulation in the WTP glass discharge riser and melter bottom as a function of glass composition, time, and temperature. When coupled with an associated operating limit (e.g., the maximum tolerable thickness of an accumulated layer of crystals), this model could then be integrated into the process control algorithms to formulate crystal tolerant high level waste (HLW) glasses targeting higher waste loadings while still meeting process related limits and melter lifetime expectancies. This report provides a review of the scaled melter testing that was completed in support of the Defense Waste Processing Facility (DWPF) melter. Testing with scaled melters provided the data to define the DWPF operating limits to avoid bulk (volume) crystallization in the un-agitated DWPF melter and provided the data to distinguish between spinels generated by K-3 refractory corrosion versus spinels that precipitated from the HLW glass melt pool. This report includes a review of the crystallization observed with the scaled melters and the full scale DWPF melters (DWPF Melter 1 and DWPF Melter 2). Examples of actual DWPF melter attainment with Melter 2 are given. The intent is to provide an overview of lessons learned, including some example data, that can be used to advance the development and implementation of an empirical model and operating limit for crystal accumulation for WTP. Operation of the first and second (current) DWPF melters has demonstrated that the strategy of using a liquidus temperature predictive model combined with a 100 °C offset from the normal melter operating temperature of 1150 °C (i.e., the predicted liquidus temperature (TL) of the glass must be 1050 °C or less) has been successful in preventing any detrimental accumulation of spinel in the DWPF melt pool, and spinel has not been

  1. F/H Area high level waste removal plan ampersand schedule as required by the Federal Facility Agreement for the Savannah River Site

    International Nuclear Information System (INIS)

    Hunter, M.A.

    1993-11-01

    The F and H-area HLW Tank Farms are one component of a larger integrated waste treatment system consisting of facilities designed for the overall processing of several radioactive waste streams resulting from nuclear material processing. Section IX.E of the SRS Federal Facility Agreement requires the DOE to submit to the EPA and SCDHEC for review and approval, a plan(s) and schedule(s) for the removal from service of waste tank systems(s)/component(s) that do not meet secondary containment standards, or that leak or have leaked. The Plan and Schedule for removal from service of these waste tanks is shown in Appendices A and B, respectively. Other portions of this package which include schedule dates are provided for information only. The SRS intends to remove systems from service as opposed to providing secondary containment for non-compliant systems. The systems that do not meet secondary containment requirements or that have leaked (as determined by tank assessment reports) include High Level Waste Tanks No. 1--24 along with corresponding ancillary equipment

  2. Application of a risk-assessment methodology to a hypothetical high-level waste repository in bedded salt

    International Nuclear Information System (INIS)

    Cranwell, R.M.; Helton, J.C.

    1982-01-01

    The Sandia/NRC Risk Assessment Methodology consists of a procedure for assessing the post-closure, long-term risk from the disposal of radioactive waste in deep geologic formations. This procedure contains: (1) methods for selecting and screening potentially disruptive events, features and processes (i.e., scenarios); (2) models for use in simulating the physical processes and estimating the potential health effects associated with the deep geologic disposal of radioactive waste (e.g., repository evolution, ground-water flow and nuclide transport, biosphere transport and human exposure, and dose commitment and dose response); and (3) probabilistic and statistical procedures for use in risk estimates and in sensitivity and uncertainty analyses. Results of the demonstration of this methodology in the analysis of a hypothetical high-level waste repository in bedded salt are presented for the following three scenarios: (1) a hydraulic communication (boreholes or shafts) connects the middle and lower sandstone aquifers allowing water to flow through the depository; (2) a hydraulic communication allows water to flow from the middle sandstone aquifer through the depository and back to middle sandstone aquifer, referred to as a U-tube scenario; and (3) withdrawl wells completed into the middle sandstone aquifer down-gradient from the depository are coupled with the U-tube of Scenario 2

  3. Laboratory-scale vitrification and leaching of Hanford high-level waste for the purpose of simulant and glass property models validation

    International Nuclear Information System (INIS)

    Morrey, E.V.; Elliott, M.L.; Tingey, J.M.

    1993-02-01

    The Hanford Waste Vitrification Plant (HWVP) is being built to process the high-level and TRU waste into canistered glass logs for disposal in a national repository. Testing programs have been established within the Project to verify process technology using simulated waste. A parallel testing program with actual radioactive waste is being performed to confirm the validity of using simulates and glass property models for waste form qualification and process testing. The first feed type to be processed by and the first to be tested on a laboratory-scale is pretreated neutralized current acid waste (NCAW). The NCAW is a neutralized high-level waste stream generated from the reprocessing of irradiated nuclear fuel in the Plutonium and Uranium Extraction (PUREX) Plant at Hanford. As part of the fuel reprocessing, the high-level waste generated in PUREX was denitrated with sugar to form current acid waste (CAW). Sodium hydroxide and sodium nitrite were added to the CAW to minimize corrosion in the tanks, thus yielding neutralized CAW. The NCAW contains small amounts of plutonium, fission products from the irradiated fuel, stainless steel corrosion products, and iron and sulfate from the ferrous sulfamate reductant used in the PUREX process. This paper will discuss the results and status of the laboratory-scale radioactive testing

  4. Modeling the corrosion of high-level waste containers: CAM-CRM interface

    International Nuclear Information System (INIS)

    Farmer, J.C.; Bedrossian, P.J.; McCright, R.D.

    1998-06-01

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological respository at Yucca Mountain, Nevada is a two-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 825, 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as A516 or Monel 400. At the present time, Alloy C-22 and A516 are favored. This publication addresses the development of models to account for corrosion of Alloy C-22 surfaces exposed directly to the Near Field Environmental (NFE), as well as to the exacerbated conditions in the CAM-CRM crevice

  5. Physical and chemical characterization of borosilicate glasses containing Hanford high-level wastes

    International Nuclear Information System (INIS)

    Kupfer, M.J.; Palmer, R.A.

    1980-10-01

    Scouting studies are being performed to develop and evaluate silicate glass forms for immobilization of Hanford high-level wastes. Detailed knowledge of the physical and chemical properties of these glasses is required to assess their suitability for long-term storage or disposal. Some key properties to be considered in selecting a glass waste form include leach resistance, resistance to radiation, microstructure (includes devitrification behavior or crystallinity), homogeneity, viscosity, electrical resistivity, mechanical ruggedness, thermal expansion, thermal conductivity, density, softening point, annealing point, strain point, glass transformation temperature, and refractive index. Other properties that are important during processing of the glass include volatilization of glass and waste components, and corrosivity of the glass on melter components. Experimental procedures used to characterize silicate waste glass forms and typical properties of selected glass compositions containing simulated Hanford sludge and residual liquid wastes are presented. A discussion of the significance and use of each measured property is also presented

  6. Waste acceptance product specifications for vitrified high-level waste forms

    International Nuclear Information System (INIS)

    Applewhite-Ramsey, A.; Sproull, J.F.

    1993-01-01

    The Nuclear Waste Policy Act of 1982 mandated that all high-level waste (HLW) be sent to a federal geologic repository for permanent disposal. DOE published the Environmental Assessment in 1982 which identified borosilicate glass as the chosen HLW form. 1 In 1985 the Department of Energy instituted a Waste Acceptance Process to assure that DWPF glass waste forms would be acceptable to such a repository. This assurance was important since production of waste forms will precede repository construction and licensing. As part of this Waste Acceptance Process, the DOE Office of Civilian Radioactive Waste Management (RW) formed the Waste Acceptance Committee (WAC). The WAC included representatives from the candidate repository sites, the waste producing sites and DOE. The WAC was responsible for developing the Waste Acceptance Preliminary Specifications (WAPS) which defined the requirements the waste forms must meet to be compatible with the candidate repository geologies

  7. Preliminary evaluation of alternative forms for immobilization of Hanford high-level wastes

    International Nuclear Information System (INIS)

    Schulz, W.W.; Beary, M.M.; Gallagher, S.A.; Higley, B.A.; Johnston, R.G.; Kupfer, M.J.; Palmer, R.A.

    1981-01-01

    Borosilicate Glass Marbles and/or monoliths were rated among the top three waste forms for immobilization of all types of Hanford high-level waste. Supergrout Concrete and Bitumen, low temperature processes, are judged to be particularly suitable for immobilization and bulk disposal of high sodium blended wastes and/or residual liquid. This preliminary assessment indicates that certain ceramic waste forms (e.g., Tailored Ceramics, Supercalcine Ceramic, and SYNROC Ceramic) are equal to or superior to Borosilicate Glass waste forms for immobilization of Hanford sludges and radionuclides removed from salt cake and residual liquid. These ceramic waste forms can be made by the Sol Gel process. Some multibarrier waste forms (e.g., Coated Ceramics, Ceramic Pellets in Metal Matrix, and Glass in Metal Matrix) are judged to be superior waste forms for immobilization of Hanford sludges and/or radionuclide concentrate

  8. Draft environmental impact statement. High-level waste repository site suitability criteria

    International Nuclear Information System (INIS)

    1978-01-01

    The purpose of HLWRSSC is to present guidelines which will help in the development of safe waste management schemes. Current regulations require solidification of all high-level waste within 5 years of their generation and transfer to a Federal waste repository within 10 years. Development of the proposed HLWRSSC is part of the overall NRC program to close the ''back end'' of the commercial LWR fuel cycle. In this document, the need for the HLWRSSC is reviewed, and the national energy policy, the need for electrical energy, and the nuclear fuel cycle are discussed. Considerations for HLWRSSC are presented, including the nature of the repository, important site-related factors, and radiological risk assessment methodology. Radiological and nonradiological environment impacts associated with the HLWRSSC are defined. Alternatives to the criteria are presented, and the cost-benefit-risk evaluation is reviewed

  9. Corrosion aspects of high-level waste disposal in salt domes

    International Nuclear Information System (INIS)

    Roerbo, K.

    1979-12-01

    In the ELSAM/ELKRAT waste management project it is planned that the high-level waste is glassified, encapsuled in canisters and finally deposited in a deep hole drilled in a salt dome. In the present report corrosion aspects of the canisters after deposition are discussed. The chemical environment will probably be a limited amount of brine coming from brine inclusions in the surrounding salt and moving up against the temperature gradient, the temperature at the canister surface being in the range of 100-150degC. The possible types of corrosion and the expected corrosion rates for a number of potential canister materials (mild steel, austenitic and ferritic stainless steels, Ni-base alloys, copper, titanium and a few combinations of materials) are discussed. Mild steel (possibly combined with an inner layer of copper or titanium) might possibly be an appropriate choice of material for the canister. (author)

  10. Seismotectonic investigations for Yucca Mountain high-level waste repository: Rationale for defining scope

    International Nuclear Information System (INIS)

    Gupta, D.C.; Blackford, M.E.

    1990-01-01

    The geologic, seismic, and engineering characteristics of the Yucca Mountain site and its environs need to be investigated in sufficient scope and detail to provide reasonable assurance that they are sufficiently well understood to permit an adequate evaluation of the proposed site for the development of a high-level waste repository. The paper examines the extent of seismotectonic investigations needed for proper evaluation of the geologic setting. At the Yucca Mountain site, a thorough understanding of tectonic phenomena such as seismicity and faulting is critical to the identification of potentially disqualifying conditions. Study of the tectonic movement, stress, or co-tectonic effects that could affect the performance of the waste-handling facilities, waste package, underground openings, shaft and borehole seals, and long-term alteration of geohydrology would be necessary. In addition, the uncertainties involved in evaluating the effect of seismotectonics on the radionuclide transport mechanism need to be thoroughly investigated. 8 refs., 1 fig

  11. Fuzzy distributions in probabilistic environmental impact assessment: application to a high-level waste repository

    International Nuclear Information System (INIS)

    Datta, D.; Joshi, M.L.

    2006-01-01

    Environmental modeling with a satisfaction levels of the end user in relation to a defined parameter coupled with imprecision that stems from the field data is a key issue. In the context of this issue success of possibility theory based on fuzzy sets has high visibility in comparison with conventional probability theory. Environmental impact assessments of a high level waste repository is focused using the new approach because the problems under consideration includes a number of qualitative uncertainties at different levels, apart from being quite complex; decision-maker's need to have a transparent assessment result that will enable him to understand underlying assumptions and to judge resulting doses. Fuzzy distributions have been tried to resolve the issues related to the safety of environment from the waste repository. Paper describes the details of fuzzy distribution, fuzzy logic and its possible application to deal the qualitative and quantitative uncertainty in connection with waste repository. (author)

  12. Alternatives for high-level waste forms, containers, and container processing systems

    International Nuclear Information System (INIS)

    Crawford, T.W.

    1995-01-01

    This study evaluates alternatives for high-level waste forms, containers, container processing systems, and onsite interim storage. Glass waste forms considered are cullet, marbles, gems, and monolithic glass. Small and large containers configured with several combinations of overpack confinement and shield casks are evaluated for these waste forms. Onsite interim storage concepts including canister storage building, bore holes, and storage pad were configured with various glass forms and canister alternatives. All favorable options include the monolithic glass production process as the waste form. Of the favorable options the unshielded 4- and 7-canister overpack options have the greatest technical assurance associated with their design concepts due to their process packaging and storage methods. These canisters are 0.68 m and 0.54 m in diameter respectively and 4.57 m tall. Life-cycle costs are not a discriminating factor in most cases, varying typically less than 15 percent

  13. Technical Exchange on Improved Design and Performance of High Level Waste Melters - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    SK Sundaram; ML Elliott; D Bickford

    1999-11-19

    SIA Radon is responsible for management of low- and intermediate-level radioactive waste (LILW) produced in Central Russia. In cooperation with Minatom organizations Radon carries out R and D programs on treatment of simulated high level waste (HLW) as well. Radon scientists deal with a study of materials for LILW, HLW, and Nuclear Power Plants (NPP) wastes immobilization, and development and testing of processes and technologies for waste treatment and disposal. Radon is mostly experienced in LILW vitrification. This experience can be carried over to HLW vitrification especially in field of melting systems. The melter chosen as a basic unit for the vitrification plant is a cold crucible. Later on Radon experience in LILW vitrification as well as our results on simulated HLW vitrification are briefly described.

  14. Radionuclide release from high level waste forms under repository conditions in clay or granite

    International Nuclear Information System (INIS)

    Godon, N.; Lanza, F.

    1990-01-01

    The behaviour of both fully active and simulated vitrified high level waste (HLW) has been studied under conditions that are likely to occur in future repositories in clay and granite. The simulated HLW was doped with Cs, Sr, Tc and the actinides and the leaching of these elements from the glass has been measured together with their concentrations in the water of the near-field and their distribution between the various components of the repository. The diffusion coefficients of several elements in Boom clay has also been measured. The results show that the concentrations of Tc and the actinides in the near-field of a repository will be very low and that the actinides will only diffuse very slowly away from the vicinity of the glass. 24 refs., 1 figs., 7 tabs

  15. Optimization of TRPO process parameters for americium extraction from high level waste

    International Nuclear Information System (INIS)

    Chen Jing; Wang Jianchen; Song Chongli

    2001-01-01

    The numerical calculations for Am multistage fractional extraction by trialkyl phosphine oxide (TRPO) were verified by a hot test. 1750L/t-U high level waste (HLW) was used as the feed to the TRPO process. The analysis used the simple objective function to minimize the total waste content in the TRPO process streams. Some process parameters were optimized after other parameters were selected. The optimal process parameters for Am extraction by TRPO are: 10 stages for extraction and 2 stages for scrubbing; a flow rate ratio of 0.931 for extraction and 4.42 for scrubbing; nitric acid concentration of 1.35 mol/L for the feed and 0.5 mol/L for the scrubbing solution. Finally, the nitric acid and Am concentration profiles in the optimal TRPO extraction process are given

  16. Plans for managing high level waste at the Savannah River Site

    International Nuclear Information System (INIS)

    Sheikh, N.A.; Salaymeh, S.R.

    1994-01-01

    Design and evaluation guidelines for DOE facilities subject to earthquake, wind/tornado, and flood have been developed. This paper describes the philosophy and procedures for the design of facilities managing high level waste, with respect to ground shaking and tornado generated missiles. The intent is to meet the probabilistic based performance goals for each category of structure, system, and component of such facilities. Described performance goals can be achieved by specifying hazard probabilities of exceedance. The seismic and tornado effects can be controlled through the level of conservatism already introduced in the design of such facilities. This will eliminate any level of risk to personnel and environment in the event of high seismic activity or high wind/tornado event. When dealing with complex processes, facilities will have to be divided into various segments and assigned different categories depending on their mission, value, or toxic/radioactive material content

  17. Integrated model of Korean spent fuel and high level waste disposal options - 16091

    International Nuclear Information System (INIS)

    Hwang, Yongsoo; Miller, Ian

    2009-01-01

    This paper describes an integrated model developed by the Korean Atomic Energy Research Institute (KAERI) to simulate options for disposal of spent nuclear fuel (SNF) and reprocessing products in South Korea. A companion paper (Hwang and Miller, 2009) describes a systems-level model of Korean options for spent nuclear fuel (SNF) management in the 21. century. The model addresses alternative design concepts for disposal of SNF of different types (Candu, PWR), high level waste, and fission products arising from a variety of alternative fuel cycle back ends. It uses the GoldSim software to simulate the engineered system, near-field and far-field geosphere, and biosphere, resulting in long-term dose predictions for a variety of receptor groups. The model's results allow direct comparison of alternative repository design concepts, and identification of key parameter uncertainties and contributors to receptor doses. (authors)

  18. Long-term high-level waste technology. Composite quarterly technical report, July-September 1980

    International Nuclear Information System (INIS)

    Cornman, W.R.

    1981-02-01

    This composite quarterly technical report summarizes work performed at participating sites to immobilize high-level radioactive wastes. The technical information included in this report is structured along the lines of the Work Breakdown Structure adopted for use in the High-Level Waste Management Technology (WBS) program. The functions and work elements of the WBS are as follows: function 1 - program management and support, which includes work elements of management and budget, environmental and safety assessments, and other support; function 2 - waste preparation, which includes in-situ storage or disposal, waste retrieval, and separation and concentration; function 3 - waste fixation with work elements of waste form development and characterization, and process and equipment development; and function 4 - final handling which includes canister development and characterization, and onsite storage or disposal

  19. Proceedings from the technical workshop on near-field performance assessment for high-level waste

    International Nuclear Information System (INIS)

    Sellin, P.; Apted, M.; Gago, J.

    1991-12-01

    This report contains the proceedings of 'Technical workshop of near-filed performance assessment for high-level waste' held in Madrid October 15-17, 1990. It includes the invited presentations and summaries of the scientific discussions. The workshop covered several topics: * post-emplacement environment, * benchmarking of computer codes, * glass release, * spent-fuel release, * radionuclide solubility, * near-field transport processes, * coupled processes in the near-field, * integrated assessments, * sensitivity analyses and validation. There was an invited presentation on each topic followed by an extensive discussion. One of the points highlighted in the closing discussion of the workshop was the need for international cooperation in the field of near-field performance assessment. The general opinion was that this was best achieved in smaller groups discussing specific questions. (au) Separate abstracts were prepared for 9 papers in this volume

  20. Swelling pressures of a potential buffer material for high-level waste repository

    International Nuclear Information System (INIS)

    Lee, Jae Owan; Cho, Won Jin; Chun, Kwan Sik

    1999-01-01

    The swelling pressure of a potential buffer material was measured and the effect of dry density, bentonite content and initial water content on the swelling pressure was investigated to provide the information for the selection of buffer material in a high-level waste repository. Swelling tests were carried out according to Box-Behnken's experimental design. Measured swelling pressures were in the wide range of 0.7 Kg/cm 2 to 190.2 Kg/cm 2 under given experimental conditions. Based upon the experimental data, a 3-factor polynomial swelling model was suggested to analyze the effect of dry density, bentonite content and initial water content on the swelling pressure. The swelling pressure increased with an increase in the dry density and bentonite content, while it decreased with increasing the initial water content and, beyond about 12 wt.% of the initial water content, levelled to nearly constant value. (author). 21 refs., 10 figs., 4 tabs

  1. Characterization of high level waste for minor actinides by chemical separation and alpha spectrometry

    International Nuclear Information System (INIS)

    Murali, M.S.; Bhattacharayya, A.; Kar, A.S.; Tomar, B.S.; Manchanda, V.K.

    2010-01-01

    Quantification of minor actinides present in of High Level Waste (HLW) solutions originating from the power reactors is important in view of management of radioactive wastes and actinide partitioning. Several methods such as ICP-MS, X-ray fluorescence methods, ICP-AES, alpha spectrometry are used in characterizing such types of wastes. As alpha spectrometry is simple and reliable, this technique has been used for the estimation of minor actinides after devising steps of separation for estimating Np and Pu present in HLW solutions of PHWR origin. Using a wealth of knowledge appropriate to the solution chemistry of actinides, the task of separation, though appears easy, it is challenging job for a radiochemist handling high-dose HLW samples, for obtaining clean alpha peaks for Np and Pu. This paper reports on the successful attempt made to quantify 241 Am, 244 Cm, Pu (239 mainly) and 237 Np present in HLW-PHWR obtained from PREFRE, Tarapur

  2. Borehole-to-borehole geophysical methods applied to investigations of high level waste repository sites

    International Nuclear Information System (INIS)

    Ramirez, A.L.

    1983-01-01

    This discussion focuses on the use of borehole to borehole geophysical measurements to detect geological discontinuities in High Level Waste (HLW) repository sites. The need for these techniques arises from: (a) the requirement that a HLW repository's characteristics and projected performance be known with a high degree of confidence; and (b) the inadequacy of other geophysical methods in mapping fractures. Probing configurations which can be used to characterize HLW sites are described. Results from experiments in which these techniques were applied to problems similar to those expected at repository sites are briefly discussed. The use of a procedure designed to reduce uncertainty associated with all geophysical exploration techniques is proposed; key components of the procedure are defined

  3. Disposal of high-level waste from nuclear power plants in Denmark. v.3

    International Nuclear Information System (INIS)

    1981-01-01

    The present report deals with material testing as part of the investigations regarding a repository for high-level waste in a salt dome. It is volume 3 of five volumes that together constitute the final report of the Danish utilities' salt dome investigations. The material testing programme chiefly comprised laboratory investigations and analyses of material samples, partly obtained from the salt dome in question and partly from the overlying geological formation. The test programme was fully completed for Eslev salt dome on Mors. Chemical analyses and mineralogical investigations were carried out in order to determine the type and extent of the impurities confined in the rock salt. Moreover, a programme was carried out to investigate the water content of the salt. Temperature conditions around a repository, the strenght and creep properties of the salt were investigated in order to determine whether drill holes, shafts and mine galleries could be constructed and maintained intact within the period of time required to complete the disposal of high-level waste. Chemical analysis were carried out in order to determine which substances are found dissolved in the water contained by the geological formation overlying the salt dome, as well as chemical/physical investigations regarding the water content, porosity, velocity of a possible leak of radioactive waste products, etc. Materials that would be introduced into a repository were studied with regard to their corrosion resistance in the salt environment. Concrete materials were investigated and characterized for their use in the final sealing of the access routes to the repository through the geological formation. (BP)

  4. Experiments on container materials for Swiss high-level waste disposal projects. Part IV

    International Nuclear Information System (INIS)

    Simpson, J.P.

    1989-12-01

    One concept for final disposal of high-level waste in switzerland consists of a repository at a depth of 1000 to 1500 m in the crystalline bedrock of Northern Switzerland. The waste will be placed in a container which will be required to function as a high integrity barrier for at least 100 years. This report is the fourth and last in the current series dealing with the evaluation of potential materials for such containers. Four materials were identified for further evaluation in the first of these reports: cast steel, nodular cast iron, copper and Ti-Code 12. This report deals with the problem of demonstrating that cast steel containers will not fail by stress corrosion cracking and with the problem of hydrogen produced by the reduction of water. The experimental results on pre-cracked specimens revealed no susceptibility of cast steel to stress corrosion cracking under model repository conditions. No crack growth was detected on compact DCB specimens exposed in aerobic and anaerobic groundwaters at 80 and 140 o C for 16-24 months. Cast steel remains a candidate material for high-level waste containers. As expected from thermodynamic considerations no hydrogen could be detected from copper immersed in model groundwaters at 50 o C. Hydrogen is evolved from corroding steel under anaerobic conditions. Hydrogen evolution due to corrosion of iron or steel in waste repositories has to be considered in any safety analysis; the amounts produced can be significant. Evidence todate suggests that both cast steel and copper are suitable container materials. Because the corrosion behaviour of both materials is sensitive to service conditions, in particular length of the aerobic phase, groundwater chemistry and temperature, further testing should be undertaken when a specific site has been identified. (author) 9 tabs., 11 figs., 25 refs

  5. 45-Day safety screen results for tank 241-U-202, push mode, cores 75 and 78

    International Nuclear Information System (INIS)

    Jo, J.

    1995-01-01

    This document is a report of the analytical results for samples collected from the radioactive wastes in Tank 241-U-202 at the Hanford Reservation. Core samples were collected from the solid wastes in the tank and underwent safety screening analyses including differential scanning calorimetry, thermogravimetric analysis, and total alpha analysis. Results indicate that no safety screening notification limits were exceeded

  6. Hanford High-Level Waste Vitrification Program at the Pacific Northwest National Laboratory: technology development - annotated bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Larson, D.E.

    1996-09-01

    This report provides a collection of annotated bibliographies for documents prepared under the Hanford High-Level Waste Vitrification (Plant) Program. The bibliographies are for documents from Fiscal Year 1983 through Fiscal Year 1995, and include work conducted at or under the direction of the Pacific Northwest National Laboratory. The bibliographies included focus on the technology developed over the specified time period for vitrifying Hanford pretreated high-level waste. The following subject areas are included: General Documentation; Program Documentation; High-Level Waste Characterization; Glass Formulation and Characterization; Feed Preparation; Radioactive Feed Preparation and Glass Properties Testing; Full-Scale Feed Preparation Testing; Equipment Materials Testing; Melter Performance Assessment and Evaluations; Liquid-Fed Ceramic Melter; Cold Crucible Melter; Stirred Melter; High-Temperature Melter; Melter Off-Gas Treatment; Vitrification Waste Treatment; Process, Product Control and Modeling; Analytical; and Canister Closure, Decontamination, and Handling

  7. Containment barrier metals for high-level waste packages in a Tuff repository

    Energy Technology Data Exchange (ETDEWEB)

    Russell, E.W.; McCright, R.D.; O`Neal, W.C.

    1983-10-12

    The Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package project is part of the US Department of Energy`s Civilian Radioactive Waste Management (CRWM) Program. The NNWSI project is working towards the development of multibarriered packages for the disposal of spent fuel and high-level waste in tuff in the unsaturated zone at Yucca Mountain at the Nevada Test Site (NTS). The final engineered barrier system design may be composed of a waste form, canister, overpack, borehole liner, packing, and the near field host rock, or some combination thereof. Lawrence Livermore National Laboratory`s (LLNL) role is to design, model, and test the waste package subsystem for the tuff repository. At the present stage of development of the nuclear waste management program at LLNL, the detailed requirements for the waste package design are not yet firmly established. In spite of these uncertainties as to the detailed package requirements, we have begun the conceptual design stage. By conceptual design, we mean design based on our best assessment of present and future regulatory requirements. We anticipate that changes will occur as the detailed requirements for waste package design are finalized. 17 references, 4 figures, 10 tables.

  8. State-of-the-art review of quality assurance techniques for vitrified high level waste

    International Nuclear Information System (INIS)

    Miller, P.L.H.

    1984-07-01

    Quality assurance is required for certain chemical and physical properties of both the molten glass pour and the solidified glass within the stainless steel container. It is also required to monitor the physical condition of the container lid weld. A review is presented of techniques which are used or which might be adapted for use in the quality assurance of vitrified high level waste. For the most part only non-intrusive methods have been considered, however, some techniques which are not strictly non-intrusive have been reviewed where a non-intrusive technique has not been identified or where there are other advantages associated with the particular technique. In order to identify suitable candidate techniques reference has been made to an extensive literature survey and experts in the fields of nuclear waste technology, glass technology, non-destructive testing, chemical analysis and remote analysis have been contacted. The opinions of manufacturers and users of specific techniques have also been sought. A summary is also given of those techniques which can most readily be applied to the problem of quality assurance for vitrified waste as well as recommendations for further research into techniques which might be adapted to suit this application. (author)

  9. Field instrumentation and testing needs for a high level waste repository

    International Nuclear Information System (INIS)

    Marti, J.; Maini, T.

    1981-03-01

    A review has been conducted of the testing and measurement needs posed by a deep geologic High Level Waste (HLW) repository in crystalline or argillaceous rocks. Siting, design, construction, operation and decommissioning of the repository have been covered, together with the planning of a Test and Demonstration Facility. Instruments and methods available have been critically assessed in their ability to fulfil the aforementioned testing and monitoring programmes. Special attention has been paid to the relation of measurements to the data needs and to the tests likely to generate such data. This assessment has concentrated on measurements of absolute rock stresses, monitoring of changes in rock stress, evaluation of the rock mass deformability, measurement of relative displacements and determination of the hydrogeologic parameters of the rock mass. Other measurements have been studied with a lesser degree of attention. The overall conclusion is that, from the instrumentation and testing points of view, present plans for a test and demonstration facility in the early nineties and a repository soon after 2000 are indeed feasible. Specific conclusions on the state-of-the-art and development needs are presented in the report. (author)

  10. High-level waste canister corrosion studies pertinent to geologic isolation

    International Nuclear Information System (INIS)

    Braithwaite, J.W.; Molecke, M.A.

    1978-12-01

    The compatibility of candidate high-level waste (HLW) canister materials with deep geologic isolation environments is addressed. Results are presented which are applicable to the following repositories or test facilities: bedded and domed salt, sub-seabed sediment, and various types of hardrock. Such studies are an essential portion of the technological basis for terminal waste management. These studies will identify HLW canister or overpack materials satisfying appropriate requirements for barrier lifetime. Mechanical properties, as well as constraints on cost and consumption of critically limited materials, are also selection criteria. Lifetime objectives range from a minimum of several years for retrievability constraints up to several hundred years for retardation of near-field interactions (e.g., waste form leaching with potential radionuclide release to the geosphere) during the period of greatest HLW thermal output. A review of present and prior applicable corrosion results is presented. However, emphasis is on the results obtained from current laboratory and in situ HLW canister/corrosion programs at Sandia Laboratories. The effects of multiple variables on corrosion susceptibility and rates are briefly discussed and some applicable data given. It is possible to provide a canister/overpack barrier which can survive geologic isolation environments for periods of several hundred years

  11. MODELING THE CORROSION OF HIGH-LEVEL WASTE CONTAINERS CAM-CRM INTERFACE

    International Nuclear Information System (INIS)

    JOSEPH C. FARMER, PETER J. BEDROSSIAN, R. DANIEL MCCRIGHT

    1998-01-01

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological repository at Yucca Mountain, Nevada is a two-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as carbon steel or Monel400. Initially, the containers will be hot and dry due to the heat generated by radioactive decay. However, the temperature will eventually drop to levels where both humid air and aqueous phase corrosion will be possible. As the outer barrier is penetrated, uniform corrosion of the CRM will be possible in exfoliated areas. The possibility for crevice formation between the CAM and CRM will also exist. In the case of either Alloy 625 or C-22, a crevice will have to form before significant penetration of the CRM can occur. Crevice corrosion of the CRMs has been well documented. Lillard and Scully have induced crevice corrosion in Alloy 625 during exposure to artificial sea water. Jones and Wilde have prepared simulated crevice solutions of FeCl 2 , NiCl 2 and CrCl 3 , and measured substantial pH suppression. Asphahani measured the dissolution rates of Alloys 625 and C-22 in such artificial crevice solutions at various temperatures. Others have observed no significant localized attack in less severe environments

  12. Leaching and mechanical properties of cabal glasses developed as matrices for immobilization high-level wastes

    International Nuclear Information System (INIS)

    Ezz-Eldin, F.M.

    2001-01-01

    This paper discusses the leaching behavior of simulated high-level-waste cabal glass (CaO-B 2 O 3 -Al 2 O 3 ) as a bulk specimen. During leach tests, the glass is immersed in either deionized water or in groundwater for up to 57 days at 70 deg. C. Based on the results, mechanisms observed with the leaching of the glass in deionized water or groundwater are discussed. Three factors, i.e., time of immersion, type of leaching solution and irradiation effect, are extensively studied. The corrosion was found to be linear with time in the limit of investigation (1-57 days) but with different rates depending on the type of solution and glass composition. Effects of γ-irradiation on the glass together with groundwater were found to decrease the glass durability. The evolution of the damage on mechanical and physical properties of the glass before and after leaching or irradiation was also discussed. The addition of waste oxide changes the properties of the glass matrix, so the influence of the guest oxides on the properties of host materials is also discussed

  13. Radiation effects in glass waste forms for high-level waste and plutonium disposal

    International Nuclear Information System (INIS)

    Weber, W.J.; Ewing, R.C.

    1997-01-01

    A key challenge in the permanent disposal of high-level waste (HLW), plutonium residues/scraps, and excess weapons plutonium in glass waste forms is the development of predictive models of long-term performance that are based on a sound scientific understanding of relevant phenomena. Radiation effects from β-decay and α-decay can impact the performance of glasses for HLW and Pu disposition through the interactions of the α-particles, β-particles, recoil nuclei, and γ-rays with the atoms in the glass. Recently, a scientific panel convened under the auspices of the DOE Council on Materials Science to assess the current state of understanding, identify important scientific issues, and recommend directions for research in the area of radiation effects in glasses for HLW and Pu disposition. The overall finding of the panel was that there is a critical lack of systematic understanding on radiation effects in glasses at the atomic, microscopic, and macroscopic levels. The current state of understanding on radiation effects in glass waste forms and critical scientific issues are presented

  14. Dose Contribution from High Level Waste Uranium and Plutonium. Revision 1

    International Nuclear Information System (INIS)

    Chandler, M.C.; Gray, P.L.; d'Entremont, P.D.; Marra, J.E.; Monahon, T.M.

    1994-01-01

    Radiological source terms for safety analyses traditionally have been curie lists of radionuclides. Converting the source term to dose values allows each radionuclide to be evaluated for its impact on dose, which is the purpose of the source term. This report is one in a series of reports establishing source terms for High Level Waste (HLW) by evaluating the dose impact of each radionuclide. These reports will be used in establishing the source terms to be used in HLW Safety Analysis Reports. The purpose of this report is to document the bounding element dose impact of uranium and plutonium in HLW. This technique (use of dose rather than curies) demonstrates vividly the relative importance of these nuclides in accident analyses. A large amount of available data permitted dose values to be established for uranium and plutonium; therefore, these two elements were evaluated independent of other nuclides. Solubility and adsorption data, available for these elements, allow bounding conditions to be established for their contribution to dose for various HLW processes

  15. Region-scale groundwater flow modelling of generic high level waste disposal sites

    International Nuclear Information System (INIS)

    Metcalfe, D.

    1996-02-01

    Regional-scale groundwater flow modelling analyses are performed on generic high level waste (HLW) disposal sites to assess the extent to which a large crystalline rock mass such as a pluton or batholith can be expected to contain and isolate HLW in terms of hydraulic considerations, for a variety of geologic and hydrogeologic conditions. The two-dimensional cross-sectional conceptual models of generic HLW disposal sites are evaluated using SWIFT III, which is a finite-difference flow and transport code. All steps leading to the final results and conclusions are incorporated in this report. The available data and information on geological and hydrogeologic conditions in plutons and batholiths are summarized. The generic conceptual models developed from this information are defined in terms of the finite difference grid, the geologic and hydrogeologic properties and the hydrologic boundary conditions used. The modelled results are described with contour maps showing the modelled head fields, groundwater flow paths and travel times and groundwater flux rates within the modelled systems. The results of the modelling analyses are used to develop general conclusions on the scales and patterns of groundwater flow in granitic plutons and batholiths. The conclusions focus on geologic and hydrogeologic characteristics that can result in favourable conditions, in terms of hydraulic considerations, for a HLW repository. (author) 43 refs., 9 tabs., 40 figs

  16. The effects of uranium oxide high-level waste on the structure of iron phosphate glasses

    International Nuclear Information System (INIS)

    Badyal, Y.

    1998-01-01

    Because of their unusually good chemical durability, iron phosphate glasses are a natural candidate for a nuclear waste disposal glass. We have studied the effects of UO 2 high-level waste on the structure of iron phosphate glasses with both neutron and high-energy x-ray diffraction using the GLAD instrument of the Intense Pulsed Neutron Source and the 1-BM bending magnet beamline of the Advanced Photon Source, respectively. The results of neutron scattering, which is mostly sensitive to correlations involving light atoms i.e. O-O, Fe-O and P-O, suggest the main structural features of the base glass are largely unaffected by the addition of UO 2 . The nearest-neighbor P-O, Fe-O and O-O peaks remain at the same position in real space and their intensities scale approximately with concentration. These findings are consistent with the earlier results of Raman scattering and EXAFS on the Fe-K edge wherein both cases the spectra remain similar to the base glass. High-energy x-ray scattering which is sensitive to correlations involving the heavier atoms and thus complements the neutron measurements, is also consistent with uranium occupying interstitial sites in the relatively undisturbed base glass structure. However, important questions remain as to the precise local structure and oxidation state of uranium in these glasses

  17. Review of metal-matrix encapsulation of solidified radioactive high-level waste

    International Nuclear Information System (INIS)

    Jardine, L.J.; Steindler, M.J.

    1978-05-01

    Literature describing previous and current work on the encapsulation of solidified high-level waste forms in a metal matrix was reviewed. Encapsulation of either stabilized calcine pellets or glass beads in alloys by casting techniques was concluded to be the most developed and direct approach to fabricating solid metal-matrix waste forms. Further characterizations of the physical and chemical properties of metal-matrix waste forms are still needed to assess the net attributes of metal-encapsulation alternatives. Steady-state heat transfer properties of waste canisters in air and water environments were calculated for four reference waste forms: (1) calcine, (2) glass monoliths, (3) metal-encapsulated calcine, and (4) metal-encapsulated glass beads. A set of criteria for the maximum allowable canister centerline and surface temperatures and heat generation rates per canister at the time of shipment to a Federal repository was assumed, and comparisons were made between canisters of these reference waste forms of the shortest time after reactor discharge that canisters could be filled and the subsequent ''interim'' storage times prior to shipment to a Federal repository for various canister diameters and waste ages. A reference conceptual flowsheet based on existing or developing technology for encapsulation of stabilized calcine pellets is discussed. Conclusions and recommendations are presented

  18. Summary Of Cold Crucible Vitrification Tests Results With Savannah River Site High Level Waste Surrogates

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovsky, Sergey; Marra, James; Lebedev, Vladimir

    2014-01-13

    The cold crucible inductive melting (CCIM) technology successfully applied for vitrification of low- and intermediate-level waste (LILW) at SIA Radon, Russia, was tested to be implemented for vitrification of high-level waste (HLW) stored at Savannah River Site, USA. Mixtures of Sludge Batch 2 (SB2) and 4 (SB4) waste surrogates and borosilicate frits as slurries were vitrified in bench- (236 mm inner diameter) and full-scale (418 mm inner diameter) cold crucibles. Various process conditions were tested and major process variables were determined. Melts were poured into 10L canisters and cooled to room temperature in air or in heat-insulated boxes by a regime similar to Canister Centerline Cooling (CCC) used at DWPF. The products with waste loading from ~40 to ~65 wt.% were investigated in details. The products contained 40 to 55 wt.% waste oxides were predominantly amorphous; at higher waste loadings (WL) spinel structure phases and nepheline were present. Normalized release values for Li, B, Na, and Si determined by PCT procedure remain lower than those from EA glass at waste loadings of up to 60 wt.%.

  19. Corrosion behaviour of container materials for geological disposal of high-level waste

    International Nuclear Information System (INIS)

    Haijtink, B.

    1986-01-01

    Within the framework of the Community R and D programme on management and storage of radioactive waste (shared cost action), a research activity is aiming at the assessment of the corrosion behaviour of potential container materials for the geological disposal of vitrified high-level waste. In a joint programme, three promising reference materials are being tested in environments representative of the three considered geological formations, clay, salt and granite. Samples of the three reference materials, Ti-0.2% Pd, Hastelloy C 4 and a low carbon steel were provided by the Commission to the participating laboratories respectively: Studiecentrum voor Kernenergie (SCK/CEN) at Mol (Belgium), Kernforschungszentrum (KfK) at Karlsruhe (Federal Republic of Germany), Commissariat a l'Energie Atomique (CEA) at Fontenay-aux-Roses (France), the Atomic Energy Research Establishment (AERE) at Harwell (United Kingdom) and the Centre National de la Recherche Scientifique (CNRS) at Vitry (France). In this report, the results obtained during the year 1984 are described

  20. Comparison of selected foreign plans and practices for spent fuel and high-level waste management

    International Nuclear Information System (INIS)

    Schneider, K.J.; Mitchell, S.J.; Lakey, L.T.; Johnson, A.B. Jr.; Hazelton, R.F.; Bradley, D.J.

    1990-04-01

    This report describes the major parameters for management of spent nuclear fuel and high-level radioactive wastes in selected foreign countries as of December 1989 and compares them with those in the United States. The foreign countries included in this study are Belgium, Canada, France, the Federal Republic of Germany, Japan, Sweden, Switzerland, and the United Kingdom. All the countries are planning for disposal of spent fuel and/or high-level wastes in deep geologic repositories. Most countries (except Canada and Sweden) plan to reprocess their spent fuel and vitrify the resultant high-level liquid wastes; in comparison, the US plans direct disposal of spent fuel. The US is planning to use a container for spent fuel as the primary engineered barrier. The US has the most developed repository concept and has one of the earliest scheduled repository startup dates. The repository environment presently being considered in the US is unique, being located in tuff above the water table. The US also has the most prescriptive regulations and performance requirements for the repository system and its components. 135 refs., 8 tabs

  1. Transportation packagings for high-level wastes and unprocessed transuranic wastes

    International Nuclear Information System (INIS)

    Wilmot, E.L.; Romesberg, L.E.

    1982-01-01

    Packagings used for nuclear waste transport are varied in size, shape, and weight because they must accommodate a wide variety of waste forms and types. However, this paper will discuss the common characteristics among the packagings in order to provide a broad understanding of packaging designs. The paper then discusses, in some detail, a design that has been under development recently at Sandia National Laboratories (SNL) for handling unprocessed, contact-handled transuranic (CHTRU) wastes as well as a cask design for defense high-level wastes (HLW). As presently conceived, the design of the transuranic package transporter (TRUPACT) calls for inner and outer boxes that are separated by a rigid polyurethane foam. The inner box has a steel frame with stainless steel surfaces; the outer box is similarly constructed except that carbon steel is used for the outside surfaces. The access to each box is through hinged doors that are sealed after loading. To meet another waste management need, a cask is being developed to transport defense HLW. The cask, which is at the preliminary design stage, is being developed by General Atomic under the direction of the TTC. The cask design relies heavily on state-of-the-art spent-fuel cask designs though it can be much simpler due to the characteristics of the HLW. A primary purpose of this paper is to show that CHTRU waste and defense HLW currently are and will be transported in packagings designed to meet the hazards of transportation that are present in general commerce

  2. Preliminary assessments for disposal of high-level waste within small islets

    International Nuclear Information System (INIS)

    Chen, F.L.; Li, S.H.; Yu, G.P.

    1991-01-01

    This paper analyzes, for the final disposal of high-level waste (HLW), the possibility of a repository sited below the fresh/saline groundwater interface within islet rock formations. Because of their relatively stable tectonics, the offshore islets of some countries (such as those of Taiwan) are worthy of being considered as potential repository sites. Before the emplacement of radwastes in such a repository, however, the mass exchange across the fresh/saline groundwater interface must be limited and the horizontal movement of advective saline ground-water must be extremely low. Theoretical equations for the location and shape of the interface are derived. When radwastes are buried in rock formations, the temperature effect of the decay heat could cause buoyant convection flow of saline groundwater upward across the groundwater interface. This could carry released radionuclides across the groundwater interface to upper formation layers where fresh groundwater flows. The radionuclides could then be carried by the fresh groundwater to the sea. Although basic HLW repository designs should eliminate the significance of this temperature effect, it is incorporated into this preliminary analysis for the purpose of conservative estimations

  3. Review of metal-matrix encapsulation of solidified radioactive high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J; Steindler, M J

    1978-05-01

    Literature describing previous and current work on the encapsulation of solidified high-level waste forms in a metal matrix was reviewed. Encapsulation of either stabilized calcine pellets or glass beads in alloys by casting techniques was concluded to be the most developed and direct approach to fabricating solid metal-matrix waste forms. Further characterizations of the physical and chemical properties of metal-matrix waste forms are still needed to assess the net attributes of metal-encapsulation alternatives. Steady-state heat transfer properties of waste canisters in air and water environments were calculated for four reference waste forms: (1) calcine, (2) glass monoliths, (3) metal-encapsulated calcine, and (4) metal-encapsulated glass beads. A set of criteria for the maximum allowable canister centerline and surface temperatures and heat generation rates per canister at the time of shipment to a Federal repository was assumed, and comparisons were made between canisters of these reference waste forms of the shortest time after reactor discharge that canisters could be filled and the subsequent ''interim'' storage times prior to shipment to a Federal repository for various canister diameters and waste ages. A reference conceptual flowsheet based on existing or developing technology for encapsulation of stabilized calcine pellets is discussed. Conclusions and recommendations are presented.

  4. West Valley demonstration project: alternative processes for solidifying the high-level wastes

    International Nuclear Information System (INIS)

    Holton, L.K.; Larson, D.E.; Partain, W.L.; Treat, R.L.

    1981-10-01

    In 1980, the US Department of Energy (DOE) established the West Valley Solidification Project as the result of legislation passed by the US Congress. The purpose of this project was to carry out a high level nuclear waste management demonstration project at the Western New York Nuclear Service Center in West Valley, New York. The DOE authorized the Pacific Northwest Laboratory (PNL), which is operated by Battelle Memorial Institute, to assess alternative processes for treatment and solidification of the WNYNSC high-level wastes. The Process Alternatives Study is the suject of this report. Two pretreatment approaches and several waste form processes were selected for evaluation in this study. The two waste treatment approaches were the salt/sludge separation process and the combined waste process. Both terminal and interim waste form processes were studied. The terminal waste form processes considered were: borosilicate glass, low-alkali glass, marbles-in-lead matrix, and crystallinolecular potential and molecular dynamics calculations of the effect are yet to be completed. Cous oxide was also investigated. The reaction is first order in nitrite ion, second order in hydrogen ion, and between zero and first order in hydroxylamine monosulfonate, depending on the concentration

  5. German Spent Nuclear Fuel Legacy: Characteristics and High-Level Waste Management Issues

    Directory of Open Access Journals (Sweden)

    A. Schwenk-Ferrero

    2013-01-01

    Full Text Available Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the paper, the assessments of German spent fuel legacy (heavy metal content and the nuclide composition of this inventory have been done. The methodology used applies advanced nuclear fuel cycle simulation techniques in order to reproduce the operation of the German nuclear power plants from 1969 till 2022. NFCSim code developed by LANL was adopted for this purpose. It was estimated that ~10,300 tonnes of unreprocessed nuclear spent fuel will be generated until the shut-down of the ultimate German reactor. This inventory will contain ~131 tonnes of plutonium, ~21 tonnes of minor actinides, and 440 tonnes of fission products. Apart from this, ca.215 tonnes of vitrified HLW will be present. As fission products and transuranium elements remain radioactive from 104 to 106 years, the characteristics of spent fuel legacy over this period are estimated, and their impacts on decay storage and final repository are discussed.

  6. Uncertainties in geologic disposal of high-level wastes - groundwater transport of radionuclides and radiological consequences

    International Nuclear Information System (INIS)

    Kocher, D.C.; Sjoreen, A.L.; Bard, C.S.

    1983-01-01

    The analysis for radionuclide transport in groundwater considers models and methods for characterizing (1) the present geologic environment and its future evolution due to natural geologic processes and to repository development and waste emplacement, (2) groundwater hydrology, (3) radionuclide geochemistry, and (4) the interactions among these phenomena. The discussion of groundwater transport focuses on the nature of the sources of uncertainty rather than on quantitative estimates of their magnitude, because of the lack of evidence that current models can provide realistic quantitative predictions of radionuclide transport in groundwater for expected repository environments. The analysis for the long-term health risk to man following releases of long-lived radionuclides to the biosphere is more quantitative and involves estimates of uncertainties in (1) radionuclide concentrations in man's exposure environment, (2) radionuclide intake by exposed individuals per unit concentration in the environment, (3) the dose per unit intake, (4) the number of exposed individuals, and (5) the health risk per unit dose. For the important long-lived radionuclides in high-level waste, uncertainties in most of the different components of a calculation of individual and collective dose per unit release appear to be no more than two or three orders of magnitude; these uncertainties are certainly much less than uncertainties in predicting groundwater transport of radionuclides between a repository and the biosphere. Several limitations in current models for predicting the health risk to man per unit release to the biosphere are discussed

  7. High level waste repository site suitability criteria. Environmental impact statement methodology

    International Nuclear Information System (INIS)

    1977-06-01

    The approach (methodology) which has been developed for the preparation of the environmental impact statement (EIS) is described. A suggested outline is presented for the High Level Waste Repository Site Suitability Criteria EIS together with a detailed description of the approach to be used in preparing the EIS. In addition, a methodology is presented by which the necessary cost/benefit/risk comparisons of alternative sets of site suitability criteria can be made. The TERA environmental research data bank, a computerized data bank which contained information on current and historical licensing activities for power plants was modified to include information on generic or programmatic EIS related issues. The content of the modified data bank was utilized to develop the EIS outline presented in this report. The report recommends that a modified matrix evaluation approach be used to make the cost/benefit/risk comparisons. The suggested matrix is designed to facilitate between criteria comparative analyses of economic, environmental, sociological and radiological risk factors. The quantitative compositing of dollar cost and benefits, environmental and sociological impacts, and radiological risks is to be performed using a semi-analytical, semi-visual procedure based on the concept of ''decision surfaces.''

  8. Health and environmental risk-related impacts of actinide burning on high-level waste disposal

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1992-05-01

    The potential health and environmental risk-related impacts of actinide burning for high-level waste disposal were evaluated. Actinide burning, also called waste partitioning-transmutation, is an advanced method for radioactive waste management based on the idea of destroying the most toxic components in the waste. It consists of two steps: (1) selective removal of the most toxic radionuclides from high-level/spent fuel waste and (2) conversion of those radionuclides into less toxic radioactive materials and/or stable elements. Risk, as used in this report, is defined as the probability of a failure times its consequence. Actinide burning has two potential health and environmental impacts on waste management. Risks and the magnitude of high-consequence repository failure scenarios are decreased by inventory reduction of the long-term radioactivity in the repository. (What does not exist cannot create risk or uncertainty.) Risk may also be reduced by the changes in the waste characteristics, resulting from selection of waste forms after processing, that are superior to spent fuel and which lower the potential of transport of radionuclides from waste form to accessible environment. There are no negative health or environmental impacts to the repository from actinide burning; however, there may be such impacts elsewhere in the fuel cycle

  9. Immobilization of defense high-level waste: an assessment of technological strategies and potential regulatory goals. Volume I

    International Nuclear Information System (INIS)

    1979-06-01

    An investigation was made of the high-level radioactive waste immobilization technology programs in the U.S. and Europe, and of the associated regulatory programs and waste management perspectives in the countries studied. Purpose was to assess the ability of those programs to satisfy DOE waste management needs and U.S. regulatory requirements. This volume includes: introduction, immobilization strategies in the context of waste isolation program needs, high-level waste management as an integrated system, regulatory goals, engineered-barrier characteristics, barrier technology, high-level waste disposal programs, analysis of HLW immobilization technology in the context of policy and regulatory requirements, and waste immobilization program option

  10. Radiological consequences of drilling intrusion into a deep repository for high level waste

    International Nuclear Information System (INIS)

    Charles, D.; McEwen, T.J.

    1991-01-01

    This report provides estimates of maximum annual individual doses to personnel for given set of scenarios associated with intrusion by drilling into a deep repository for high level waste (HLW) and recovery to the surface of contaminated material. Geological and engineering judgment has been used to formulate the manner of this intrusion and the parameters which would determine the resulting doses to field workers and to others who might be involved subsequently in procedures carried out in a geotechnical laboratory. Calculations have been performed to quantify the magnitudes of doses resulting from drilling into the HLW spent fuel matrix, the bentonite buffer and the adjacent rock. Doses have been calculated in the basis of the potential for external irradiation, inhalation of suspended contamination, and inadvertent ingestion of small amounts of contaminated material. The results show that intrusion doses into spent fuel within a few decades of disposal could be very high, in the range about 10 to 100 Sv. At this stage all three exposure pathways make a significant contribution to the total dose. Beyond about 100 y, as 137 Cs decays away, the contribution from external irradiation falls off sharply. However the doses from ingestion and inhalation have only fallen by about a factor of 5 after 10 3 y, and the total dose still in the range from about 2 to 25 Sv. Apart form the direct consequences of intrusion discussed above, drilling might influence safety in other ways. Thus, a borehole might modify the groundwater flow system around the waste, and provide a direct permeable flow path to the near surface; it might results in the introduction of drilling fluids which modify groundwater chemistry; it might results in the early failure of a fuel canister, leading to earlier release to groundwater; or, in the extreme, some combination of all three. (au)

  11. Idaho National Engineering Laboratory High-Level Waste Roadmap. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    The Idaho National Engineering Laboratory (INEL) High-Level Waste (HLW) Roadmap takes a strategic look at the entire HLW life-cycle starting with generation, through interim storage, treatment and processing, transportation, and on to final disposal. The roadmap is an issue-based planning approach that compares ``where we are now`` to ``where we want and need to be.`` The INEL has been effectively managing HLW for the last 30 years. Calcining operations are continuing to turn liquid HLW into a more manageable form. Although this document recognizes problems concerning HLW at the INEL, there is no imminent risk to the public or environment. By analyzing the INEL current business operations, pertinent laws and regulations, and committed milestones, the INEL HLW Roadmap has identified eight key issues existing at the INEL that must be resolved in order to reach long-term objectives. These issues are as follows: A. The US Department of Energy (DOE) needs a consistent policy for HLW generation, handling, treatment, storage, and disposal. B. The capability for final disposal of HLW does not exist. C. Adequate processes have not been developed or implemented for immobilization and disposal of INEL HLW. D. HLW storage at the INEL is not adequate in terms of capacity and regulatory requirements. E. Waste streams are generated with limited consideration for waste minimization. F. HLW is not adequately characterized for disposal nor, in some cases, for storage. G. Research and development of all process options for INEL HLW treatment and disposal are not being adequately pursued due to resource limitations. H. HLW transportation methods are not selected or implemented. A root-cause analysis uncovered the underlying causes of each of these issues.

  12. Physical modeling of joule heated ceramic glass melters for high level waste immobilization

    International Nuclear Information System (INIS)

    Quigley, M.S.; Kreid, D.K.

    1979-03-01

    This study developed physical modeling techniques and apparatus suitable for experimental analysis of joule heated ceramic glass melters designed for immobilizing high level waste. The physical modeling experiments can give qualitative insight into the design and operation of prototype furnaces and, if properly verified with prototype data, the physical models could be used for quantitative analysis of specific furnaces. Based on evaluation of the results of this study, it is recommended that the following actions and investigations be undertaken: It was not shown that the isothermal boundary conditions imposed by this study established prototypic heat losses through the boundaries of the model. Prototype wall temperatures and heat fluxes should be measured to provide better verification of the accuracy of the physical model. The VECTRA computer code is a two-dimensional analytical model. Physical model runs which are isothermal in the Y direction should be made to provide two-dimensional data for more direct comparison to the VECTRA predictions. The ability of the physical model to accurately predict prototype operating conditions should be proven before the model can become a reliable design tool. This will require significantly more prototype operating and glass property data than were available at the time of this study. A complete set of measurements covering power input, heat balances, wall temperatures, glass temperatures, and glass properties should be attempted for at least one prototype run. The information could be used to verify both physical and analytical models. Particle settling and/or sludge buildup should be studied directly by observing the accumulation of the appropriate size and density particles during feeding in the physical model. New designs should be formulated and modeled to minimize the potential problems with melter operation identifed by this study

  13. Physical modeling of joule heated ceramic glass melters for high level waste immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Quigley, M.S.; Kreid, D.K.

    1979-03-01

    This study developed physical modeling techniques and apparatus suitable for experimental analysis of joule heated ceramic glass melters designed for immobilizing high level waste. The physical modeling experiments can give qualitative insight into the design and operation of prototype furnaces and, if properly verified with prototype data, the physical models could be used for quantitative analysis of specific furnaces. Based on evaluation of the results of this study, it is recommended that the following actions and investigations be undertaken: It was not shown that the isothermal boundary conditions imposed by this study established prototypic heat losses through the boundaries of the model. Prototype wall temperatures and heat fluxes should be measured to provide better verification of the accuracy of the physical model. The VECTRA computer code is a two-dimensional analytical model. Physical model runs which are isothermal in the Y direction should be made to provide two-dimensional data for more direct comparison to the VECTRA predictions. The ability of the physical model to accurately predict prototype operating conditions should be proven before the model can become a reliable design tool. This will require significantly more prototype operating and glass property data than were available at the time of this study. A complete set of measurements covering power input, heat balances, wall temperatures, glass temperatures, and glass properties should be attempted for at least one prototype run. The information could be used to verify both physical and analytical models. Particle settling and/or sludge buildup should be studied directly by observing the accumulation of the appropriate size and density particles during feeding in the physical model. New designs should be formulated and modeled to minimize the potential problems with melter operation identifed by this study.

  14. Final disposal of high-level waste from nuclear power plants

    International Nuclear Information System (INIS)

    1979-05-01

    A report on the possibilities of high-level radioactive waste disposal is presented by a working group appointed by the Danish Academy of Technical Sciences which consists of specialists in nuclear power, corrosion, health physics, chemistry, geology, physics, and zoology. If nuclear power is introduced in Denmark five nuclear power plants with a total installed power of about 5000 MW are expected to be established. Dependent on the handling procedures the volume of the high-level wastes will be between 10 and 50 cubic meters yearly. As the amount of wastes that would be produced in Danish nuclear power plants will be small compared to the amount from other countries, it is natural to try to solve a Danish waste problem through an arrangement on waste disposal with a foreign plant. Also, for economic and technical reasons this solution would be the best, and the entering into a satisfactory agreement should therefore be given high priority. However, it is not possible to know for sure that such an agreement could be reached, and it is therefore necessary to evaluate the possibilities for geologic waste disposal in Denmark. It is concluded that without doubt several geologic formations in the Danish deeper underground are suitable for waste disposal. A main problem is to find a formation which can prove to be isolated from the biosphere for many millennia. It is recommended that geological formations be studied in detail and that some locations suitable for waste disposal should be selected in order that they can be reserved for this purpose. (B.P.)

  15. ENHANCED DOE HIGH LEVEL WASTE MELTER THROUGHPUT STUDIES: SRNL GLASS SELECTION STRATEGY

    Energy Technology Data Exchange (ETDEWEB)

    Raszewski, F; Tommy Edwards, T; David Peeler, D

    2008-01-23

    The Department of Energy has authorized a team of glass formulation and processing experts at the Savannah River National Laboratory (SRNL), the Pacific Northwest National Laboratory (PNNL), and the Vitreous State Laboratory (VSL) at Catholic University of America to develop a systematic approach to increase high level waste melter throughput (by increasing waste loading with minimal or positive impacts on melt rate). This task is aimed at proof-of-principle testing and the development of tools to improve waste loading and melt rate, which will lead to higher waste throughput. Four specific tasks have been proposed to meet these objectives (for details, see WSRC-STI-2007-00483): (1) Integration and Oversight, (2) Crystal Accumulation Modeling (led by PNNL)/Higher Waste Loading Glasses (led by SRNL), (3) Melt Rate Evaluation and Modeling, and (4) Melter Scale Demonstrations. Task 2, Crystal Accumulation Modeling/Higher Waste Loading Glasses is the focus of this report. The objective of this study is to provide supplemental data to support the possible use of alternative melter technologies and/or implementation of alternative process control models or strategies to target higher waste loadings (WLs) for the Defense Waste Processing Facility (DWPF)--ultimately leading to higher waste throughputs and a reduced mission life. The glass selection strategy discussed in this report was developed to gain insight into specific technical issues that could limit or compromise the ability of glass formulation efforts to target higher WLs for future sludge batches at the Savannah River Site (SRS). These technical issues include Al-dissolution, higher TiO{sub 2} limits and homogeneity issues for coupled-operations, Al{sub 2}O{sub 3} solubility, and nepheline formation. To address these technical issues, a test matrix of 28 glass compositions has been developed based on 5 different sludge projections for future processing. The glasses will be fabricated and characterized based on

  16. Methods of calculating the post-closure performance of high-level waste repositories

    International Nuclear Information System (INIS)

    Ross, B.

    1989-02-01

    This report is intended as an overview of post-closure performance assessment methods for high-level radioactive waste repositories and is designed to give the reader a broad sense of the state of the art of this technology. As described here, ''the state of the art'' includes only what has been reported in report, journal, and conference proceedings literature through August 1987. There is a very large literature on the performance of high-level waste repositories. In order to make a review of this breadth manageable, its scope must be carefully defined. The essential principle followed is that only methods of calculating the long-term performance of waste repositories are described. The report is organized to reflect, in a generalized way, the logical order to steps that would be taken in a typical performance assessment. Chapter 2 describes ways of identifying scenarios and estimating their probabilities. Chapter 3 presents models used to determine the physical and chemical environment of a repository, including models of heat transfer, radiation, geochemistry, rock mechanics, brine migration, radiation effects on chemistry, and coupled processes. The next two chapters address the performance of specific barriers to release of radioactivity. Chapter 4 treats engineered barriers, including containers, waste forms, backfills around waste packages, shaft and borehole seals, and repository design features. Chapter 5 discusses natural barriers, including ground water systems and stability of salt formations. The final chapters address optics of general applicability to performance assessment models. Methods of sensitivity and uncertainty analysis are described in Chapter 6, and natural analogues of repositories are treated in Chapter 7. 473 refs., 19 figs., 2 tabs

  17. Potential Application Of Radionuclide Scaling Factors To High Level Waste Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Reboul, S. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2013-09-30

    Production sources, radiological properties, relative solubilities in waste, and laboratory analysis techniques for the forty-five radionuclides identified in Hanford's Waste Treatment and Immobilization Plant (WTP) Feed Acceptance Data Quality Objectives (DQO) document are addressed in this report. Based on Savannah River Site (SRS) experience and waste characteristics, thirteen of the radionuclides are judged to be candidates for potential scaling in High Level Waste (HLW) based on the concentrations of other radionuclides as determined through laboratory measurements. The thirteen radionuclides conducive to potential scaling are: Ni-59, Zr-93, Nb-93m, Cd-113m, Sn-121m, Sn-126, Cs-135, Sm-151, Ra-226, Ra-228, Ac-227, Pa-231, and Th-229. The ability to scale radionuclides is useful from two primary perspectives: 1) it provides a means of checking the radionuclide concentrations that have been determined by laboratory analysis; and 2) it provides a means of estimating radionuclide concentrations in the absence of a laboratory analysis technique or when a complex laboratory analysis technique fails. Along with the rationale for identifying and applying the potential scaling factors, this report also provides examples of using the scaling factors to estimate concentrations of radionuclides in current SRS waste and into the future. Also included in the report are examples of independent laboratory analysis techniques that can be used to check results of key radionuclide analyses. Effective utilization of radionuclide scaling factors requires understanding of the applicable production sources and the chemistry of the waste. As such, the potential scaling approaches identified in this report should be assessed from the perspective of the Hanford waste before reaching a decision regarding WTP applicability.

  18. Methods of calculating the post-closure performance of high-level waste repositories

    Energy Technology Data Exchange (ETDEWEB)

    Ross, B. (ed.)

    1989-02-01

    This report is intended as an overview of post-closure performance assessment methods for high-level radioactive waste repositories and is designed to give the reader a broad sense of the state of the art of this technology. As described here, ''the state of the art'' includes only what has been reported in report, journal, and conference proceedings literature through August 1987. There is a very large literature on the performance of high-level waste repositories. In order to make a review of this breadth manageable, its scope must be carefully defined. The essential principle followed is that only methods of calculating the long-term performance of waste repositories are described. The report is organized to reflect, in a generalized way, the logical order to steps that would be taken in a typical performance assessment. Chapter 2 describes ways of identifying scenarios and estimating their probabilities. Chapter 3 presents models used to determine the physical and chemical environment of a repository, including models of heat transfer, radiation, geochemistry, rock mechanics, brine migration, radiation effects on chemistry, and coupled processes. The next two chapters address the performance of specific barriers to release of radioactivity. Chapter 4 treats engineered barriers, including containers, waste forms, backfills around waste packages, shaft and borehole seals, and repository design features. Chapter 5 discusses natural barriers, including ground water systems and stability of salt formations. The final chapters address optics of general applicability to performance assessment models. Methods of sensitivity and uncertainty analysis are described in Chapter 6, and natural analogues of repositories are treated in Chapter 7. 473 refs., 19 figs., 2 tabs.

  19. Effect of Feed Melting, Temperature History, and Minor Component Addition on Spinel Crystallization in High-Level Waste Glass

    Czech Academy of Sciences Publication Activity Database

    Izák, Pavel; Hrma, P.; Arey, B. W.; Plaisted, T. J.

    2001-01-01

    Roč. 289, 1-3 (2001), s. 17-29 ISSN 0022-3093 Grant - others:DOE(US) DE/06/76RL01830 Keywords : feed melting * crystalization * high-level waste glass Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 1.363, year: 2001

  20. Geochemical Processes Controlling Migration of High Level Wastes in Hanford's Vadose Zone

    International Nuclear Information System (INIS)

    Zachara, John M.; Serne, R. Jeffrey; Freshley, Mark D.; Mann, Frederick M.; Anderson, Frank J.; Wood, Marcus I.; Jones, Thomas E.; Myers, David A.

    2007-01-01

    High level nuclear wastes (HLW) from Hanford's plutonium reprocessing are stored in massive, buried, single-shell tanks in eighteen tank farms. The wastes were initially hot because of radioactive decay, and many exhibited extreme chemical character in terms of pH, salinity, and radionuclide concentration. At present, 67 of the 149 single shell tanks are suspected to have released over 1.9 million L of tank waste to the vadose zone, with most leak events occurring between 1950 and 1975. Boreholes have been placed through the largest vadose zone plumes to define the extent of contaminant migration, and to develop conceptual models of processes governing the transformation, retardation, and overall transport of tank waste residuals. Laboratory studies with sediments so collected have shown that ion exchange, precipitation and dissolution, and surface complexation reactions have occurred between the HLW and subsurface sediments moderating their chemical character, and retarding the migration of select contaminants. Processes suspected to facilitate the far-field migration of immobile radionuclides including stable aqueous complex formation and mobile colloids were found to be potentially operative, but unlikely to occur in the field, with the exception of cyanide-facilitated migration of 60Co. Fission product oxyanions are the most mobile of tank waste constituents because their adsorption is suppressed by large concentrations of waste anions; the vadose zone clay fraction is negative in surface charge; and, unlike Cr, their reduced forms are unstable in oxidizing environments. Reaction/process-based transport modeling is beginning to be used for predictions of future contaminant mobility and plume evolution

  1. R and D for actinide partitioning and recovery of valuables from high level waste using radiotracers

    International Nuclear Information System (INIS)

    Manchanda, V.K.

    2006-01-01

    In the context of growing world population with rapidly increasing energy needs and the threat of global warming due to CO 2 emission (caused by fossil fuel burning), the nuclear energy may be an attractive option particularly in the developing countries. Recycling of fuel is a unique feature of nuclear power technology which makes it a favourable choice with respect to conservation of energy resources. Steady growth of global fuel reprocessing activities (6000 tHM/annum) implies a vital role of separation science in developing efficient procedures for the separation and purification of actinides and in devising safe procedures for the management of nuclear waste arising at different stages of the PUREX process. High Level Waste (HLW) comprising of the concentrate of the raffinate of the co-extraction cycle (with over 95% of the total radioactivity produced in the burn up process in reactor) need to be isolated from the biosphere. There is a consensus among the waste management technologists that the safest route to achieve this, is to deposit it in a stable geological formation after it's immobilization in suitable glass/Synroc matrix. It ensures that any risk from exposure due to accidental intervention or natural disturbance is minimized. Risk perception is essentially due to the large radiological toxicity associated with alpha emitters like 237 Np, 241 Am, 243 Am and 245 Cm. Isotopes of Pu (left unrecovered) present in HLW also contribute towards radiological toxicity. In view of the high cost involved and the need for continuous surveillance, several countries are considering modifying their reprocessing schemes to partition (isolate) long-lived actinides from HLW. Since the volume of the actinide oxides (which retain major fraction of the radio toxicity of HLW) is significantly lower as compared to the other metal oxides present in HLW, such an approach is expected to reduce the cost of immobilization as well as of disposal (in geological repository) and

  2. Nucleation and crystal growth behavior of nepheline in simulated high-level waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Amoroso, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Mcclane, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-09-26

    The Savannah River National Laboratory (SRNL) has been tasked with supporting glass formulation development and process control strategies in key technical areas, relevant to the Department of Energy’s Office of River Protection (DOE-ORP) and related to high-level waste (HLW) vitrification at the Waste Treatment and Immobilization Plant (WTP). Of specific interest is the development of predictive models for crystallization of nepheline (NaAlSiO4) in HLW glasses formulated at high alumina concentrations. This report summarizes recent progress by researchers at SRNL towards developing a predicative tool for quantifying nepheline crystallization in HLW glass canisters using laboratory experiments. In this work, differential scanning calorimetry (DSC) was used to obtain the temperature regions over which nucleation and growth of nepheline occur in three simulated HLW glasses - two glasses representative of WTP projections and one glass representative of the Defense Waste Processing Facility (DWPF) product. The DWPF glass, which has been studied previously, was chosen as a reference composition and for comparison purposes. Complementary quantitative X-ray diffraction (XRD) and optical microscopy confirmed the validity of the methodology to determine nucleation and growth behavior as a function of temperature. The nepheline crystallization growth region was determined to generally extend from ~ 500 to >850 °C, with the maximum growth rates occurring between 600 and 700 °C. For select WTP glass compositions (high Al2O3 and B2O3), the nucleation range extended from ~ 450 to 600 °C, with the maximum nucleation rates occurring at ~ 530 °C. For the DWPF glass composition, the nucleation range extended from ~ 450 to 750 °C with the maximum nucleation rate occurring at ~ 640 °C. The nepheline growth at the peak temperature, as determined by XRD, was between 35 - 75 wt.% /hour. A maximum nepheline growth rate of ~ 0.1 mm/hour at 700 °C was measured for the DWPF

  3. Geochemical modelling of bentonite porewater in high-level waste repositories

    Science.gov (United States)

    Wersin, Paul

    2003-03-01

    The description of the geochemical properties of the bentonite backfill that serves as engineered barrier for nuclear repositories is a central issue for perfomance assessment since these play a large role in determining the fate of contaminants released from the waste. In this study the porewater chemistry of bentonite was assessed with a thermodynamic modelling approach that includes ion exchange, surface complexation and mineral equilibrium reactions. The focus was to identify the geochemical reactions controlling the major ion chemistry and acid-base properties and to explore parameter uncertainties specifically at high compaction degrees. First, the adequacy of the approach was tested with two distinct surface complexation models by describing recent experimental data performed at highly varying solid/liquid ratios and ionic strengths. The results indicate adequate prediction of the entire experimental data set. Second, the modelling was extended to repository conditions, taking as an example the current Swiss concept for high-level waste where the compacted bentonite backfill is surrounded by argillaceous rock. The main reactions controlling major ion chemistry were found to be calcite equilibrium and concurrent Na-Ca exchange reactions and de-protonation of functional surface groups. Third, a sensitivity analysis of the main model parameters was performed. The results thereof indicate a remarkable robustness of the model with regard to parameter uncertainties. The bentonite system is characterised by a large acid-base buffering capacity which leads to stable pH-conditions. The uncertainty in pH was found to be mainly induced by the pCO 2 of the surrounding host rock. The results of a simple diffusion-reaction model indicate only minor changes of porewater composition with time, which is primarily due to the geochemical similarities of the bentonite and the argillaceous host rock. Overall, the results show the usefulness of simple thermodynamic models to

  4. Preliminary flowsheet for the conversion of Hanford high-level waste to glass

    International Nuclear Information System (INIS)

    Beary, M.M.; Chick, L.A.; Ely, P.C.; Gott, S.A.

    1977-06-01

    The flowsheets describe a process for converting waste removed from the Hanford underground waste tanks to more immobile form. The process involves a chemical separation of the radionuclides from industrial chemicals, and then making glass from the resulting small volume of highly radioactive waste. Removal of Sr, actinides, cesium, and technetium is discussed

  5. Alternatives to High-Level Waste Vitrification: The Need for Common Sense

    International Nuclear Information System (INIS)

    Bell, Jimmy T.

    2000-01-01

    The competition for government funding for remediation of defense wastes (and for other legitimate government functions) is intensifying as the United States moves toward a balanced national budget. Determining waste remediation priorities for the use of available tax dollars will likely depend on established international agreements and on the real risks posed to human health.Remediation of the U.S. Department of Energy (DOE) high-level radioactive tank wastes has been described as the most important priority in the DOE system. The proposed tank waste remediation at three DOE sites will include retrieval of the wastes from the aging storage tanks, immobilization of the wastes, and safe disposal of the processed waste. Vitrification, the current immobilization technology chosen by DOE, is very costly. The U.S. Congress and the American people may not be aware that the present cost of preparing just 1 m 3 of processed waste product at the Savannah River Site is ∼$2 million! In a smaller waste remediation project at the West Valley Site, similar waste treatment is costing >$2 million/m 3 of waste product. Privatization efforts at the Hanford Site are now estimated to cost >$4 million/m 3 of waste product. Even at the lowest current cost of $2 million/m 3 of HLW glass product, the total estimated costs for remediating the tank wastes at the three DOE sites of Savannah River, Hanford, and Idaho Falls is $75 billion.Whether our nation can afford treatment costs of this magnitude and whether Congress will be willing to appropriate these huge sums for waste vitrification when alternative technologies can provide safe disposal at considerably lower cost are questions that need to be addressed. The hazard levels posed by the DOE tank wastes do not warrant high priority in comparison to the hazards of other defense wastes. Unless DOE selects a lower-cost technology for tank waste remediation, such efforts are likely to continue in a holding pattern, with little actually

  6. Chemical Speciation of Sr, Am and Cm in high Level Waste: predictive modeling of phase partitioning during tank processing

    International Nuclear Information System (INIS)

    Choppin, Gregory R.

    2006-01-01

    During this contract period, a number of papers were published. The papers prior to this report have been reported in earlier annual reports. This final report covers the 2005 and 2006 publications which have been published as well as the last few which have been submitted, but are still under review for acceptance for publication. The titles and abstracts of the papers are presented in section A, and the full published papers in Section B

  7. Chemical Speciation of Sr, Am and Cm in high Level Waste: predictive modeling of phase partitioning during tank processing

    Energy Technology Data Exchange (ETDEWEB)

    Choppin, Gregory R.

    2006-12-20

    During this contract period, a number of papers were published. The papers prior to this report have been reported in earlier annual reports. This final report covers the 2005 & 2006 publications which have been published as well as the last few which have been submitted, but are still under review for acceptance for publication. The titles and abstracts of the papers are presented in section A, and the full published papers in Section B.

  8. High level waste storage tank farms/242-A evaporator standards/requirements identification document phase 1 assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Biebesheimer, E., Westinghouse Hanford Co.

    1996-09-30

    This document, the Standards/Requirements Identification Document (S/RID) Phase I Assessment Report for the subject facility, represents the results of an Administrative Assessment to determine whether S/RID requirements are fully addressed by existing policies, plans or procedures. It contains; compliance status, remedial actions, and an implementing manuals report linking S/RID elements to requirement source to implementing manual and section.

  9. Advanced waste form and Melter development for treatment of troublesome high-level wastes

    Energy Technology Data Exchange (ETDEWEB)

    Marra, James [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kim, Dong -Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maio, Vincent [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these “troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (also with high Al2O3 concentrations). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group. An extended duration CCIM melter test was conducted on an AZ-101 waste simulant using the CCIM platform at the Idaho National Laboratory (INL). The melter was continually operated for approximately 80 hours demonstrating that the AZ-101 high waste loading glass composition could be readily processed using the CCIM technology. The resulting glass was close to the targeted composition and exhibited excellent durability in both

  10. Operating experience during high-level waste vitrification at the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Valenti, P.J.; Elliott, D.I.

    1999-01-01

    This report provides a summary of operational experiences, component and system performance, and lessons learned associated with the operation of the Vitrification Facility (VF) at the West Valley Demonstration Project (WVDP). The VF was designed to convert stored high-level radioactive waste (HLW) into a stable waste form (borosilicate glass) suitable for disposal in a federal repository. Following successful completion on nonradioactive test, HLW processing began in July 1995. Completion of Phase 1 of HLW processing was reached on 10 June 1998 and represented the processing of 9.32 million curies of cesium-137 (Cs-137) and strontium-90 (Sr-90) to fill 211 canisters with over 436,000 kilograms of glass. With approximately 85% of the total estimated curie content removed from underground waste storage tanks during Phase 1, subsequent operations will focus on removal of tank heel wastes

  11. Operating experience during high-level waste vitrification at the West Valley Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    Valenti, P.J.; Elliott, D.I.

    1999-01-01

    This report provides a summary of operational experiences, component and system performance, and lessons learned associated with the operation of the Vitrification Facility (VF) at the West Valley Demonstration Project (WVDP). The VF was designed to convert stored high-level radioactive waste (HLW) into a stable waste form (borosilicate glass) suitable for disposal in a federal repository. Following successful completion on nonradioactive test, HLW processing began in July 1995. Completion of Phase 1 of HLW processing was reached on 10 June 1998 and represented the processing of 9.32 million curies of cesium-137 (Cs-137) and strontium-90 (Sr-90) to fill 211 canisters with over 436,000 kilograms of glass. With approximately 85% of the total estimated curie content removed from underground waste storage tanks during Phase 1, subsequent operations will focus on removal of tank heel wastes.

  12. Evaluation of the safety of vitrified high level waste shipments from the UK to continental Europe by sea. Annex 2

    International Nuclear Information System (INIS)

    Lange, F.; Fett, H.J.; Hoermann, E.; Roewekamp, M.; Cheshire, R.; Elston, B.; Slawson, G.; Raffestin, D.; Schneider, T.; Armingaud, F.; Laurent, B.

    2001-01-01

    The return of vitrified high level waste arising from the reprocessing of spent nuclear fuel at Sellafield to continental Europe, e.g. Germany, will start around the end of the century. The shipment of the specific flasks will include transportation via the Irish Sea, the English Channel and the North Sea with ships of the Pacific Nuclear Transport Limited (PNTL) classified to the INF 3 standard. The assessment approach is to analyse the severity and the frequency of mechanical impacts, fires and explosions with the potential to affect the package. The results show that there is a high safety margin due to the special safety features of the INF 3 ships compared to conventional ships. The remaining accident probability for a trans-port of vitrified high level waste from UK to the continent is very low. No realistic severe accident scenarios that could seriously affect the flasks and could lead to a radioactivity re-lease have been identified. (author)

  13. Progress report on safety research of high-level waste management for the period April 1987 to March 1988

    International Nuclear Information System (INIS)

    Nakamura, Haruto; Tashiro, Shingo

    1988-10-01

    Researches on high-level waste management at the High Level Waste Management Laboratory and the Waste Safety Testing Facility Operation Division of the Japan Atomic Energy Research Institute in the fiscal year of 1987 are reviewed in the three sections of the report. The topics are as follows: 1) On performance and durability of waste forms and engineered barrier materials, accelerated alpha radiation stability of glass form and Synroc has been investigated and stress corrosion cracking of canister materials was examined under simulated conditions. 2) Sorption of 237 Np on granite samples and behavior of iron during weathering of granites were studied with respect to safety evaluation for geological disposal. 3) Actual waste was transported from the Tokai Reprocessing Plant and hot operation using the actual waste was initiated at WASTEF. (author)

  14. Canister design concepts for disposal of spent fuel and high level waste

    International Nuclear Information System (INIS)

    Patel, R.; Punshon, C.; Nicholas, J.; Bastid, P.; Zhou, R.; Schneider, C.; Bagshaw, N.; Howse, D.; Hutchinson, E.; Asano, R.; King, S.

    2012-10-01

    As part of its long-term plans for development of a repository for spent fuel (SF) and high level waste (HLW), Nagra is exploring various options for the selection of materials and design concepts for disposal canisters. The selection of suitable canister options is driven by a series of requirements, one of the most important of which is providing a minimum 1000 year lifetime without breach of containment. One candidate material is carbon steel, because of its relatively low corrosion rate under repository conditions and because of the advanced state of overall technical maturity related to construction and fabrication. Other materials and design options are being pursued in parallel studies. The objective of the present study was to develop conceptual designs for carbon steel SF and HLW canisters along with supporting justification. The design process and outcomes result in design concepts that deal with all key aspects of canister fabrication, welding and inspection, short-term performance (handling and emplacement) and long-term performance (corrosion and structural behaviour after disposal). A further objective of the study is to use the design process to identify the future work that is required to develop detailed designs. The development of canister designs began with the elaboration of a number of design requirements that are derived from the need to satisfy the long-term safety requirements and the operational safety requirements (robustness needed for safe handling during emplacement and potential retrieval). It has been assumed based on radiation shielding calculations that the radiation dose rate at the canister surfaces will be at a level that prohibits manual handling, and therefore a hot cell and remote handling will be needed for filling the canisters and for final welding operations. The most important canister requirements were structured hierarchically and set in the context of an overall design methodology. Conceptual designs for SF canisters

  15. Preliminary Technology Maturation Plan for Immobilization of High-Level Waste in Glass Ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Vienna, John D.; Crum, Jarrod V.; Sevigny, Gary J.; Smith, G L.

    2012-09-30

    A technology maturation plan (TMP) was developed for immobilization of high-level waste (HLW) raffinate in a glass ceramics waste form using a cold-crucible induction melter (CCIM). The TMP was prepared by the following process: 1) define the reference process and boundaries of the technology being matured, 2) evaluate the technology elements and identify the critical technology elements (CTE), 3) identify the technology readiness level (TRL) of each of the CTE’s using the DOE G 413.3-4, 4) describe the development and demonstration activities required to advance the TRLs to 4 and 6 in order, and 5) prepare a preliminary plan to conduct the development and demonstration. Results of the technology readiness assessment identified five CTE’s and found relatively low TRL’s for each of them: • Mixing, sampling, and analysis of waste slurry and melter feed: TRL-1 • Feeding, melting, and pouring: TRL-1 • Glass ceramic formulation: TRL-1 • Canister cooling and crystallization: TRL-1 • Canister decontamination: TRL-4 Although the TRL’s are low for most of these CTE’s (TRL-1), the effort required to advance them to higher values. The activities required to advance the TRL’s are listed below: • Complete this TMP • Perform a preliminary engineering study • Characterize, estimate, and simulate waste to be treated • Laboratory scale glass ceramic testing • Melter and off-gas testing with simulants • Test the mixing, sampling, and analyses • Canister testing • Decontamination system testing • Issue a requirements document • Issue a risk management document • Complete preliminary design • Integrated pilot testing • Issue a waste compliance plan A preliminary schedule and budget were developed to complete these activities as summarized in the following table (assuming 2012 dollars). TRL Budget Year MSA FMP GCF CCC CD Overall $M 2012 1 1 1 1 4 1 0.3 2013 2 2 1 1 4 1 1.3 2014 2 3 1 1 4 1 1.8 2015 2 3 2 2 4 2 2.6 2016 2 3 2 2 4 2 4

  16. RESULTS OF THE FY09 ENHANCED DOE HIGH LEVEL WASTE MELTER THROUGHPUT STUDIES AT SRNL

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, F.; Edwards, T.

    2010-06-23

    High-level waste (HLW) throughput (i.e., the amount of waste processed per unit time) is a function of two critical parameters: waste loading (WL) and melt rate. For the Waste Treatment and Immobilization Plant (WTP) at the Hanford Site and the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). The objective of this task is to develop data, assess property models, and refine or develop the necessary models to support increased WL of HLW at SRS. It is a continuation of the studies initiated in FY07, but is under the specific guidance of a Task Change Request (TCR)/Work Authorization received from DOE headquarters (Project Number RV071301). Using the data generated in FY07, FY08 and historical data, two test matrices (60 glasses total) were developed at the Savannah River National Laboratory (SRNL) in order to generate data in broader compositional regions. These glasses were fabricated and characterized using chemical composition analysis, X-ray Diffraction (XRD), viscosity, liquidus temperature (TL) measurement and durability as defined by the Product Consistency Test (PCT). The results of this study are summarized below: (1) In general, the current durability model predicts the durabilities of higher waste loading glasses quite well. A few of the glasses exhibited poorer durability than predicted. (2) Some of the glasses exhibited anomalous behavior with respect to durability (normalized leachate for boron (NL [B])). The quenched samples of FY09EM21-02, -07 and -21 contained no nepheline or other wasteform affecting crystals, but have unacceptable NL [B] values (> 10 g/L). The ccc sample of FY09EM21-07 has a NL [B] value that is more than one half the value of the quenched sample. These glasses also have lower concentrations of Al{sub 2}O{sub 3} and SiO{sub 2}. (3) Five of the ccc samples (EM-13, -14, -15, -29 and

  17. Heat transfer analysis of the geologic disposal of spent fuel and high-level waste storage canisters

    International Nuclear Information System (INIS)

    Allen, G.K.

    1980-08-01

    Near-field temperatures resulting from the storage of high-level waste canisters and spent unreprocessed fuel assembly canisters in geologic formations were determined. Preliminary design of the repository was modeled for a heat transfer computer code, HEATING5, which used the Crank-Nicolson finite difference method to evaluate transient heat transfer. The heat transfer system was evaluated with several two- and three-dimensional models which transfer heat by a combination of conduction, natural convention, and radiation. Physical properties of the materials in the model were based upon experimental values for the various geologic formations. The effects of canister spacing, fuel age, and use of an overpack were studied for the analysis of the spent fuel canisters; salt, granite, and basalt were considered as the storage media for spent fuel canisters. The effects of canister diameter and use of an overpack were studied for the analysis of the high-level waste canisters; salt was considered as the only storage media for high-level waste canisters. Results of the studies on spent fuel assembly canisters showed that the canisters could be stored in salt formations with a maximum heat loading of 134 kw/acre without exceeding the temperature limits set for salt stability. The use of an overpack had little effect on the peak canister temperatures. When the total heat load per acre decreased, the peak temperatures reached in the geologic formations decreased; however, the time to reach the peak temperatures increased. Results of the studies on high-level waste canisters showed that an increased canister diameter will increase the canister interior temperatures considerably; at a constant areal heat loading, a 381 mm diameter canister reached almost a 50 0 C higher temperature than a 305 mm diameter canister. An overpacked canister caused almost a 30 0 C temperature rise in either case

  18. High Level Waste Feed Delivery AZ-101 Batch Transfer to the Private Contractor Transfer and Mixing Process Improvements

    International Nuclear Information System (INIS)

    DUNCAN, G.P.

    2000-01-01

    The primary purpose of this business case is to provide Operations and Maintenance with a detailed transfer process review for the first High Level Waste (HLW) feed delivery to the Privatization Contractor (PC), AZ-101 batch transfer to PC. The Team was chartered to identify improvements that could be implemented in the field. A significant penalty can be invoked for not providing the quality, quantity, or timely delivery of HLW feed to the PC

  19. 40 Years of Experience of NIRAS / Belgoprocess on the Interim Storage of Low, Intermediate and High Level Waste

    International Nuclear Information System (INIS)

    Braeckeveldt, Marnix; Ghys, Bart

    2016-01-01

    Conclusion: • ONDRAF/NIRAS and Belgoprocess have gained over time an extended experience on the interim storage of Low-Intermediate and High level waste. • An systematic inspection strategy was developed in order the verify the conformity of the different waste-packages and corrective measures were taken to guarantee safe storage conditions. • From 2022 , ONDRAF/NIRAS will operate a surface disposal facility for LLW

  20. Evaluation of health and safety impacts of defense high-level waste in geologic repositories. Draft 1

    Energy Technology Data Exchange (ETDEWEB)

    Kocher, D.C.; Witherspoon, J.P.; Smith, E.D.

    1984-11-01

    This report is concerned with evaluating the health and safety aspects of defense waste disposal during both the operational and the post-closure phase of a repository. In each case, the evaluation includes three aspects: (1) an identification and discussion of the various factors which are expected to influence the health and safety impacts of the different disposal options for defense high-level waste, (2) an identification of the general assumptions which were used in estimating potential health and safety effects and a selection of appropriate models for estimating the health and safety impacts of the various disposal options, and (3) an analysis of the health and safety impacts for each disposal option for defense high-level waste. This report describes our initial results in these areas. Based on the evaluations presented in this report, our initial conclusion is that the potential health and safety impacts are not likely to vary significantly among the different disposal options that might be chosen for defense high-level waste, primarily because of the need to meet standards in all cases. The differences in estimated health and safety aspects for different options are in all cases much smaller than the uncertainties which will be associated with realistic estimates of these impacts.

  1. 10 CFR Appendix D to Part 2 - Schedule for the Proceeding on Consideration of Construction Authorization for a High-Level Waste...

    Science.gov (United States)

    2010-01-01

    ... Construction Authorization for a High-Level Waste Geologic Repository. D Appendix D to Part 2 Energy NUCLEAR.... D Appendix D to Part 2—Schedule for the Proceeding on Consideration of Construction Authorization for a High-Level Waste Geologic Repository. Day Regulation (10 CFR) Action 0 2.101(f)(8), 2.105(a)(5...

  2. EFFECTIVENESS OF USING DILUTE OXALIC ACID TO DISSOLVE HIGH LEVEL WASTE IRON BASED SLUDGE SIMULANT

    International Nuclear Information System (INIS)

    Ketusky, E

    2008-01-01

    At the Savannah River Site (SRS), near Aiken South Carolina, there is a crucial need to remove residual quantities of highly radioactive iron-based sludge from large select underground storage tanks (e.g., 19,000 liters of sludge per tank), in order to support tank closure. The use of oxalic acid is planned to dissolve the residual sludge, hence, helping in the removal. Based on rigorous testing, primarily using 4 and 8 wt% oxalic acid solutions, it was concluded that the more concentrated the acid, the greater the amount of residual sludge that would be dissolved; hence, a baseline technology on using 8 wt% oxalic acid was developed. In stark contrast to the baseline technology, reports from other industries suggest that the dissolution will most effectively occur at 1 wt% oxalic acid (i.e., maintaining the pH near 2). The driver for using less oxalic acid is that less (i.e., moles) would decrease the severity of the downstream impacts (i.e., required oxalate solids removal efforts). To determine the initial feasibility of using 1 wt% acid to dissolve > 90% of the sludge solids, about 19,000 liters of representative sludge was modeled using about 530,000 liters of 0 to 8 wt% oxalic acid solutions. With the chemical thermodynamic equilibrium based software results showing that 1 wt% oxalic acid could theoretically work, simulant dissolution testing was initiated. For the dissolution testing, existing simulant was obtained, and an approximate 20 liter test rig was built. Multiple batch dissolutions of both wet and air-dried simulant were performed. Overall, the testing showed that dilute oxalic acid dissolved a greater fraction of the stimulant and resulted in a significantly larger acid effectiveness (i.e., grams of sludge dissolved/mole of acid) than the baseline technology. With the potential effectiveness confirmed via simulant testing, additional testing, including radioactive sludge testing, is planned

  3. Preliminary conceptual design of a geological disposal system for high-level wastes from the pyroprocessing of PWR spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui-Joo, E-mail: hjchoi@kaeri.re.kr [Korea Atomic Energy Research Institute, 1045 Daeduk-Daero, Yuseong, Daejon 305-353 (Korea, Republic of); Lee, Minsoo; Lee, Jong Youl [Korea Atomic Energy Research Institute, 1045 Daeduk-Daero, Yuseong, Daejon 305-353 (Korea, Republic of)

    2011-08-15

    Highlights: > A geological disposal system consists of disposal overpacks, a buffer, and a deposition hole or a disposal tunnel for high-level wastes from a pyroprocessing of PWR spent fuels is proposed. The amount and characteristics of high-level wastes are analyzed based on the material balance of pyroprocessing. > Four kinds of deposition methods, two horizontal and two vertical, are proposed. Thermal design is carried out with ABAQUS program. The spacing between the disposal modules is determined for the peak temperature in buffer not to exceed 100 deg. C. > The effect of the double-layered buffer is compared with the traditional single-layered buffer in terms of disposal density. Also, the effect of cooling time (aging) is illustrated. > All the thermal calculations are represented by comparing the disposal area of PWR spent fuels with the same cooling time. - Abstract: The inventories of spent fuels are linearly dependent on the production of electricity generated by nuclear energy. Pyroprocessing of PWR spent fuels is one of promising technologies which can reduce the volume of spent fuels remarkably. The properties of high-level wastes from the pyroprocessing are totally different from those of spent fuels. A geological disposal system is proposed for the high-level wastes from pyroprocessing of spent fuels. The amount and characteristics of high-level wastes are analyzed based on the material balance of pyroprocessing. Around 665 kg of monazite ceramic wastes are expected from the pyroprocessing of 10 MtU of PWR spent fuels. Decay heat from monazite ceramic wastes is calculated using the ORIGEN-ARP program. Disposal modules consisting of storage cans, overpacks, and a deposition hole or a disposal tunnel are proposed. Four kinds of deposition methods are proposed. Thermal design is carried out with ABAQUS program and geological data obtained from the KAERI Underground Research Tunnel. Through the thermal analysis, the spacing between the disposal modules

  4. Processing of High Level Waste: Spectroscopic Characterization of Redox Reactions in Supercritical Water - Final Report

    International Nuclear Information System (INIS)

    Arrington, C. A. Jr.

    2000-01-01

    Current efforts are focused on the oxidative dissolution of chromium compounds found in Hanford tank waste sludge. Samples of chromium oxides and hydroxides with varying degrees of hydration are being characterized using Raman, FTIR, and XPS spectroscopic techniques. Kinetics of oxidation reactions at subcritical and supercritical temperatures are being followed by Raman spectroscopy using a high temperature stainless steel cell with diamond windows. In these reactions both hydrogen peroxide and nitrate anions are used as the oxidizing species with Cr(III) compounds and organic compounds as reducing agents

  5. Systems engineering management and implementation plan for Project W-464, immobilized high-level waste storage

    International Nuclear Information System (INIS)

    Wecks, M.D.

    1998-01-01

    The Systems Engineering Management and Implementation Plan (SEMIP) for TWRS Project W-46 describes the project implementation of the Tank Waste Remediation System Systems Engineering Management Plan. (TWRS SEMP), Rev. 1. The SEMIP outlines systems engineering (SE) products and processes to be used by the project for technical baseline development. A formal graded approach is used to determine the products necessary for requirements, design, and operational baseline completion. SE management processes are defined, and roles and responsibilities for management processes and major technical baseline elements are documented

  6. Production of solidified high level wastes: a cost comparison of solidification processes

    International Nuclear Information System (INIS)

    1977-06-01

    Differential cost estimates of the annual operating and maintenance costs and the capital costs for five HLW Waste Solidification Alternates were developed. The annual operating and maintenance cost estimates included the cost of labor, consumables, utilities, shipping casks, shipping and disposal at a federal repository. The capital cost included the cost of the component, installation and building. The differential cost estimates do not include equipment and facilities which are either shared with the reprocessing facility or are common between all of the alternates. Total annual cost differential between the five waste form alternates is summarized in tabular form. The Borosilicate Glass Alternate has the lowest total annual cost. The other alternates have higher costs which range from $6.6 M to $7.4 M per year higher than the Glass alternate with the Supercalcine being the highest cost at $7.4 M per year differential. The major items in the cost estimates are then disposal costs in the operating cost estimates and the HLW Storage Tanks in the capital cost estimates. The Supercalcine Multibarrier Alternate ships 180 canisters per year more than the other alternates and consequently has a significantly higher operating cost. However, off-setting this the Supercalcine Multibarrier Alternate does not require HLW Storage Tanks for decay because of the high heat conductivity of this product and correspondingly the capital cost for this alternate is significantly lower than the other alternates. The radiological risk values are correlated with the cost evaluation normalized to cost ($)/MWe-yr

  7. Critical review of welding technology for canisters for disposal of spent fuel and high level waste

    International Nuclear Information System (INIS)

    Pike, S.; Allen, C.; Punshon, C.; Threadgill, P.; Gallegillo, M.; Holmes, B.; Nicholas, J.

    2010-03-01

    Nagra is the Swiss national cooperative for the disposal of radioactive waste and is responsible for final disposal of all types of waste produced in Switzerland, which are partitioned into two repository types, one for spent fuel (SF), vitrified high-level waste (HLW) and long-lived intermediate level waste and one for low and intermediate level waste. In the general licences applied for these repositories, documentation has to show that long-term safety can be ensured and that factors for the construction, operation, and closure of the facility have been considered. Nagra has commissioned TWI to carry out a critical review of welding technologies for the sealing of HLW and SF canisters made of carbon steel. In conjunction with a material selection report, the information gained will be used as a preliminary step to provide input to developing design concepts for the canisters. The features to be considered are: a) Suitability of techniques for thickness of weld required; b) Suitability for remote operation, maintenance and set-up; c) Welding speed, weld quality, tolerances and cost; d) Effect of welding process on parent materials properties including microstructure corrosion resistance, distortion and residual stress; e) Potential post-weld treatments to reduce residual stress and enhance corrosion resistance; f) Suitability of inspection techniques for the weld thickness required; g) Impact of welding techniques on the canister design and material selection; h) Critique of emerging technologies which may be suitable in the future. The review of potential welding technologies began with a feasibility study carried out by TWI experts, where the unsuitable processes were rejected. For the remaining processes attention was focused on previous applications for the material and thickness suggested, and especially on safety critical applications such as applied in the nuclear and pressure vessel industry. Once the relevant information was gathered each process was

  8. SOLUBILITY OF URANIUM AND PLUTONIUM IN ALKALINE SAVANNAH RIVER SITE HIGH LEVEL WASTE SOLUTIONS

    Energy Technology Data Exchange (ETDEWEB)

    King, W.; Hobbs, D.; Wilmarth, B.; Edwards, T.

    2010-03-10

    Five actual Savannah River Site tank waste samples and three chemically-modified samples were tested to determine solubility limits for uranium and plutonium over a one year time period. Observed final uranium concentrations ranged from 7 mg U/L to 4.5 g U/L. Final plutonium concentrations ranged from 4 {micro}g Pu/L to 12 mg Pu/L. Actinide carbonate complexation is believed to result in the dramatic solubility increases observed for one sample over long time periods. Clarkeite, NaUO{sub 2}(O)OH {center_dot} H{sub 2}O, was found to be the dominant uranium solid phase in equilibrium with the waste supernate in most cases.

  9. Response to Environmental Policy Institute report on Savannah River Plant high-level waste management

    International Nuclear Information System (INIS)

    1986-01-01

    In a recently published report entitled ''Deadly Crop in the Tank Farm,'' the Environmental Policy Institute (EPI) presented its opinions on the hazards associated with waste management practices at SRP. The EPI's allegations were based on selected published Department of Energy and Du Pont reports and on raw data from the unpublished 200-Area Fault Tree Data Bank that it obtained in 1983. Professional staff at SRP have reviewed the report in detail and have provided responses in this document to all significant EPI statements and recommendations. The responses are grouped into five major categories: Waste Management Operations -- Past and Present, Accidents and Risks, Worker Exposure and Cancer Epidemiology, Groundwater Contamination, and Long-Term Waste Management. An overview of the responses is provided in the Summary, and the detailed responses are presented in the body of the report. 55 refs., 1 fig., 1 tab

  10. Modeling of the Savannah River Site High Level Waste Evaporator Systems

    International Nuclear Information System (INIS)

    Hang, T.

    2003-01-01

    Three evaporators are used to reduce the volume of waste in the waste tank farm at the Savannah River Site (SRS). Evaporators are crucial operation in the SRS waste processing and management system. Using the Aspen Custom Modeler(TM) (ACM) software package marketed by Aspen Technology, Inc., the evaporator dynamic flowsheet models have been constructed to simulate the behavior of the evaporator systems. The evaporator models are used to assist operations and planning. The models account for the basic arrangement and flowpath for the evaporators: (1) Feed system, (2) Concentrate system, (3) Overheads system, and (4) Steam system. This paper provides a detailed description of the model development and presents the result of a typical simulation scenario

  11. Borosilicate glass as a matrix for immobilization of SRP high-level waste

    International Nuclear Information System (INIS)

    Wicks, G.G.

    1980-01-01

    Approximately 22 million gallons of high-level radioactive defense waste are currently being stored in large underground tanks located on the Savannah River Plant (SRP) site in Aiken, South Carolina. One option now being considered for long-term management of this waste involves removing the waste from the tanks, chemically processing the waste, and immobilizing the potentially harmful radionuclides in the waste into a borosilicate glass matrix. The technology for producing waste glass forms is well developed and has been demonstrated on various scales using simulated as well as radioactive SRP waste. Recently, full-scale prototypical equipment has been made operational at SRP. This includes both a joule-heated ceramic melter and an in-can melter. These melters are a part of an integrated vitrification system which is under evaluation and includes a spray calciner, direct liquid feed apparatus, and various elements of an off-gas system. Two of the most important properties of the waste glass are mechanical integrity and leachability. Programs are in progress at SRL aimed at minimizing thermally induced cracking by carefully controlling cooling cycles and using ceramic liners or coatings. The leachability of SRP waste glass has been studied under many different conditions and consistently found to be low. For example, the leachability of actual SRP waste glass was found to be 10 -6 to 10 -5 g/(cm 2 )(day) initially and decreasing to 10 -9 to 10 -8 g/(cm 2 )(day) after 100 days. Waste glass is also being studied under anticipated storage conditions. In brine at 90 0 C, the leachability is about 5 x 10 -8 g/(cm 2 )(day) after 60 days. The effects of other geological media including granite, basalt, shale, and tuff are also being studied as part of the multibarrier isolation system

  12. Evaluation of Hanford high level waste vitrification chemistry for an NCAW simulant -- FY 1994: Potential exothermic reactions in the presence of formic acid, glycolic acid, and oxalic acid

    Energy Technology Data Exchange (ETDEWEB)

    Sills, J.A.

    1995-07-01

    A potential for an uncontrollable exothermic reaction between nitrate and organic salts during preparation of a high level waste melter feed has been identified. In order to examine this potential more closely, the thermal behavior of simulated neutralized current acid waste (NCAW) treated with various organic reductants was studied. Differential scanning calorimetry (DSC) measurements were collected on simulated waste samples and their supernates treated with organics. Organic reductants used were formic acid, glycolic acid, and oxalic acid. For comparison, samples of untreated simulant and untreated simulant with added noble metals were tested. When heated, untreated simulant samples both with and without noble metals showed no exothermic behavior. All of the treated waste simulant samples showed exothermic behavior. Onset temperatures of exothermic reactions were 120 C to 210 C. Many onset temperatures, particularly those for formic acid treated samples, are well below 181 C, the estimated maximum steam coil temperature (considered to be a worst case maximum temperature for chemical process tank contents). The enthalpies of the reactions were {minus}180 {times} 10{sup {minus}3} J/Kg supernate ({minus}181 J/g) for the oxalic acid treated simulant supernate to {minus}1,150 {times} 10{sup {minus}3} J/Kg supernate ({minus}1,153 J/g) for the formic acid treated simulant supernate.

  13. Evaluation of Hanford high level waste vitrification chemistry for an NCAW simulant -- FY 1994: Potential exothermic reactions in the presence of formic acid, glycolic acid, and oxalic acid

    International Nuclear Information System (INIS)

    Sills, J.A.

    1995-07-01

    A potential for an uncontrollable exothermic reaction between nitrate and organic salts during preparation of a high level waste melter feed has been identified. In order to examine this potential more closely, the thermal behavior of simulated neutralized current acid waste (NCAW) treated with various organic reductants was studied. Differential scanning calorimetry (DSC) measurements were collected on simulated waste samples and their supernates treated with organics. Organic reductants used were formic acid, glycolic acid, and oxalic acid. For comparison, samples of untreated simulant and untreated simulant with added noble metals were tested. When heated, untreated simulant samples both with and without noble metals showed no exothermic behavior. All of the treated waste simulant samples showed exothermic behavior. Onset temperatures of exothermic reactions were 120 C to 210 C. Many onset temperatures, particularly those for formic acid treated samples, are well below 181 C, the estimated maximum steam coil temperature (considered to be a worst case maximum temperature for chemical process tank contents). The enthalpies of the reactions were -180 x 10 -3 J/Kg supernate (-181 J/g) for the oxalic acid treated simulant supernate to -1,150 x 10 -3 J/Kg supernate (-1,153 J/g) for the formic acid treated simulant supernate

  14. Success story: Nuclear production, fuel cycle and high level waste management in France: a contribution to long term energy sustainability

    International Nuclear Information System (INIS)

    Debes, Michel

    2006-01-01

    Nuclear energy, with no CO 2 emission, is a prominent contributor to energy sustainability for long term electricity production, but within a set of major conditions for its continuation and development: - Maintaining both a high level of safety, in accordance with international state of art, and the overall competitiveness of nuclear energy in the long term; - Pursuing the development of credible, efficient and economic solutions for a good management of high level waste issue, while meeting the long term institutional and financial responsibilities; - Maintaining the nuclear option fully open, with long term operation of the existing NPPs, preparation of advanced future LWR and future design (Generation IV); - Providing continued effective proliferation resistance of nuclear energy systems; - Preserving future option for energy resource. This 'success story' will focus on the situation in France for nuclear production, spent fuel and high level waste management and mid term prospective. The paper has the following structure: Introduction; Nuclear industry in France; The current fuel cycle; The current studies for future back end options; Conclusion. In conclusion one underlines that nuclear generation plays a major role in France, as a safe, clean, economic and independent energy resource, and will continue in the long term, while preparing for advanced reactors (EPR and GEN4). The closed fuel cycle strategy implemented in France, involving reprocessing of spent fuel and MOX recycling, has gained an extensive experience feedback and has reached an industrial and mature stage. It brings a robust answer to HLW treatment, with vitrification of high level waste within glass canister, while preserving energy resource for long term and meeting cost effectiveness imperative with existing facilities operated in the long run

  15. Development of integrated mechanistically-based degradation-mode models for performance assessment of high-level waste containers

    International Nuclear Information System (INIS)

    Farmer, J. C.

    1998-01-01

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological repository at Yucca Mountain, Nevada is a two-tayer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 825, 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as A516 Gr 55 or Monel 400. At the present time, Alloy C- 22 and A516 Gr 55 are favored

  16. Effect of buffer thickness on the retardation of radionuclide release from the high-level waste repository

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Won Jin; Lee, Jae Owan; Kang, Chul Hyung; Han, Kyung Won

    2000-12-01

    The radionuclide release from buffer in the high-level waste repository to the surrounding host rock was assessed, and the effect of the radial buffer thickness on the release rate was analyzed. The total release rates decrease sharply with increasing radial buffer thickness up to 0.25 m, and decrease moderately at the buffer thickness between 0.25 m and 0.5 m. But increasing the radial buffer thickness beyond 0.5 m has little effect in reducing radionuclide release. Therefore a radial buffer thickness between 0.25 m and 0.5 m is sufficient based on the viewpoint of radionuclide retention.

  17. The relation of equity issues to risk perceptions and socioeconomic impacts of a high level waste repository

    International Nuclear Information System (INIS)

    Colglazier, E.W.

    1989-01-01

    This paper discusses how equity concerns are an important component of the debater on controversial public policy decisions such as those regarding high-level waste disposal. In science policy, concepts of fairness need to be applied to what counts as evidence, as well as to process (means) and outcomes (ends). A science policy decision that involves technologies perceived as risky may have a better chance of enduring over time if all the key parties feel that they have been fairly treated. One task of risk communication is to engage in a dialogue with affected parties and the public about what is fair in terms of process, evidence, and outcomes

  18. Technology development for long-term management of Hanford high-level waste. Quarterly report, January 1976--March 1976

    Energy Technology Data Exchange (ETDEWEB)

    Kounts, J S [ed.

    1976-06-01

    This document is the seventh in a series of quarterly reports on work in support of long-term management of Hanford high-level wastes. The work reported here was performed during the period January through March 1976. The specific topics discussed are grouped into the subject areas of: Storage System Integrity and Engineered Improvements; Waste Retrieval; Waste Immobilization and Storage; Contaminated Equipment Volume Reduction; and Analytical Methods Development. Previous reports were published as ARH-ST-110 A through D and ARH-ST-132 A and B.

  19. Development of integraded mechanistically-based degradation-mode models for performance assessment of high-level waste containers

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J. C., LLNL

    1998-06-01

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological repository at Yucca Mountain, Nevada is a two-tayer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 825, 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as A516 Gr 55 or Monel 400. At the present time, Alloy C- 22 and A516 Gr 55 are favored.

  20. Prediction of geological and mechanical processes while disposing of high-level waste (HLW) into the earth crust

    International Nuclear Information System (INIS)

    Kedrovsky, O.L.; Morozov, V.N.

    1992-01-01

    Prediction of geological and mechanical processes while disposing of high-level waste of atomic industry into the earth crust is the fundamental base for ecological risk assessment (possible consequences) while developing repository designs. The subject of this paper is the analytical estimate of possibilities of rock fracturing mechanisms to predict isolation properties loss by massif beginning from crystal lattice of minerals up to large fracture disturbances under conditions of long-term influence of pressure, temperature, and radiation. To solve the problem possibilities of kinetic

  1. Comparison of potential health and safety impacts of different disposal options for defense high-level wastes

    International Nuclear Information System (INIS)

    Kocher, D.C.; Smith, E.D.; Witherspoon, J.P.

    1984-01-01

    A comparative assessment has been performed of the potential long- and short-term health and safety impacts of different disposal options for defense high-level wastes. Conservative models and assumptions were used. The assessment suggests that considerations of health and safety will not be significant in choosing among disposal options, primarily because of the need to meet stringent standards in all cases. Rather, the ease and cost of assuring compliance of a particular disposal option with health and safety standards may be a more important factor. 11 references

  2. A Review and Analysis of European Industrial Experience in Handling LWR Spent Fuel and Vitrified High-Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    Blomeke, J.O.

    2001-07-10

    The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performance of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States.

  3. High-level waste borosilicate glass a compendium of corrosion characteristics. Volume 1

    International Nuclear Information System (INIS)

    Cunnane, J.C.

    1994-03-01

    Current plans call for the United States Department of Energy (DOE) to start up facilities for vitrification of high-level radioactive waste (HLW) stored in tanks at the Savannah River Site, Aiken, South Carolina, in 1995; West Valley Demonstration Project, West Valley, New York, in 1996; and at the Hanford Site, Richland, Washington, after the year 2000. The product from these facilities will be canistered HLW borosilicate glass, which will be stored, transported, and eventually disposed of in a geologic repository. The behavior of this glass waste product, under the range of likely service conditions, is the subject of considerable scientific and public interest. Over the past few decades, a large body of scientific information on borosilicate waste glass has been generated worldwide. The intent of this document is to consolidate information pertaining to our current understanding of waste glass corrosion behavior and radionuclide release. The objective, scope, and organization of the document are discussed in Section 1.1, and an overview of borosilicate glass corrosion is provided in Section 1.2. The history of glass as a waste form and the international experience with waste glass are summarized in Sections 1.3 and 1.4, respectively

  4. High-level waste borosilicate glass a compendium of corrosion characteristics. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Cunnane, J.C. [comp.; Bates, J.K.; Bradley, C.R. [Argonne National Lab., IL (United States)

    1994-03-01

    Current plans call for the United States Department of Energy (DOE) to start up facilities for vitrification of high-level radioactive waste (HLW) stored in tanks at the Savannah River Site, Aiken, South Carolina, in 1995; West Valley Demonstration Project, West Valley, New York, in 1996; and at the Hanford Site, Richland, Washington, after the year 2000. The product from these facilities will be canistered HLW borosilicate glass, which will be stored, transported, and eventually disposed of in a geologic repository. The behavior of this glass waste product, under the range of likely service conditions, is the subject of considerable scientific and public interest. Over the past few decades, a large body of scientific information on borosilicate waste glass has been generated worldwide. The intent of this document is to consolidate information pertaining to our current understanding of waste glass corrosion behavior and radionuclide release. The objective, scope, and organization of the document are discussed in Section 1.1, and an overview of borosilicate glass corrosion is provided in Section 1.2. The history of glass as a waste form and the international experience with waste glass are summarized in Sections 1.3 and 1.4, respectively.

  5. Proposed method for assigning metric tons of heavy metal values to defense high-level waste forms to be disposed of in a geologic repository

    International Nuclear Information System (INIS)

    1987-08-01

    A proposed method is described for assigning an equivalent metric ton heavy metal (eMTHM) value to defense high-level waste forms to be disposed of in a geologic repository. This method for establishing a curie equivalency between defense high-level waste and irradiated commercial fuel is based on the ratio of defense fuel exposure to the typical commercial fuel exposure, MWd/MTHM. application of this technique to defense high-level wastes is described. Additionally, this proposed technique is compared to several alternate calculations for eMTHM. 15 refs., 2 figs., 10 tabs

  6. Long-term performance of container materials for high-level waste: [Technical report, April 1982-August 1987

    International Nuclear Information System (INIS)

    Beavers, J.A.; Thompson, N.G.; Markworth, A.J.; Cialone, H.J.; Majumdar, B.S.; McCoy, J.K.

    1987-11-01

    This report describes the results of experimental and analytical studies of high-level waste container degradation. Corrosion and hydrogen embrittlement tests were conducted on selected materials to identify environmental and metallurgical factors that promote material degradation, especially stress-corrosion cracking. A major emphasis on overpack materials focused on cast and wrought low-carbon steels. Results of the corrosion work show that, to more completely identify potential failure modes, exposure environments must be further defined. Predictions of pitting rates based on models utilizing nonreactive walls may lead to rejection of carbon steel as a viable overpack material when, on the basis of performance, it may perform satisfactorily. Hydrogen embrittlement was shown to be promoted in regions of microstructural change such as the weld heat-affected zone. These findings show that hydrogen embrittlement is important to container integrity. A small portion of this task was devoted to studying the possible internal corrosion of the canister. It was found that Type 304L stainless steel will likely contain high-level waste glass for the retrieval period and probably the thermal period. Modeling studies focused on general corrosion and pitting corrosion, with the models being extended to account for more realistic conditions. Results show that pit wall reactivity is an important consideration in predicting corrosion rates. 102 refs., 132 figs., 57 tabs

  7. Assessment, evaluation, and testing of technologies for environmental restoration, decontamination, and decommissioning and high level waste management. Progress report

    Energy Technology Data Exchange (ETDEWEB)

    Uzochukwu, G.A.

    1997-12-31

    Nuclear and commercial non-nuclear technologies that have the potential of meeting the environmental restoration, decontamination and decommissioning, and high-level waste management objectives are being assessed and evaluated. A detailed comparison of innovative technologies available will be performed to determine the safest and most economical technology for meeting these objectives. Information derived from this effort will be matched with the multi-objectives of the environmental restoration, decontamination and decommissioning, and high-level waste management effort to ensure that the best, most economical, and the safest technologies are used in decision making at USDOE-SRS. Technology-related variables will be developed and the resulting data formatted and computerized for multimedia systems. The multimedia system will be made available to technology developers and evaluators to ensure that the best, most economical, and the safest technologies are used in decision making at USDOE-SRS. Technology-related variables will be developed and the resulting data formatted and computerized for multimedia systems. The multimedia system will be made available to technology developers and evaluators to ensure that the safest and most economical technologies are developed for use at SRS and other DOE sites.

  8. Assessment, evaluation, and testing of technologies for environmental restoration, decontamination, and decommissioning and high level waste management. Progress report

    International Nuclear Information System (INIS)

    Uzochukwu, G.A.

    1997-01-01

    Nuclear and commercial non-nuclear technologies that have the potential of meeting the environmental restoration, decontamination and decommissioning, and high-level waste management objectives are being assessed and evaluated. A detailed comparison of innovative technologies available will be performed to determine the safest and most economical technology for meeting these objectives. Information derived from this effort will be matched with the multi-objectives of the environmental restoration, decontamination and decommissioning, and high-level waste management effort to ensure that the best, most economical, and the safest technologies are used in decision making at USDOE-SRS. Technology-related variables will be developed and the resulting data formatted and computerized for multimedia systems. The multimedia system will be made available to technology developers and evaluators to ensure that the best, most economical, and the safest technologies are used in decision making at USDOE-SRS. Technology-related variables will be developed and the resulting data formatted and computerized for multimedia systems. The multimedia system will be made available to technology developers and evaluators to ensure that the safest and most economical technologies are developed for use at SRS and other DOE sites

  9. Survey and analysis of the domestic technology level for the concept development of high level waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Chang Sun; Kim, Byung Su; Song, Jae Hyok [Seoul National University, Seoul (Korea); Park, Kwang Hon; Hwang, Ju Ho; Park, Sung Hyun; Lee, Jae Min [Kyunghee University, Seoul (Korea); Han, Joung Sang; Kim, Ku Young [Yonsei University, Seoul (Korea); Lee, Jae Ki; Chang, Jae Kwon [Hangyang University, Seoul (Korea)

    1998-09-01

    The objectives of this study are the analysis of the status of HLW disposal technology and the investigation of the domestic technology level. The study has taken two years to complete with the participation of forty five researchers. The study was mainly carried out through means of literature surveys, collection of related data, visits to research institutes, and meetings with experts in the specific fields. During the first year of this project, the International Symposium on the Concept Development of the High Level Waste Disposal System was held in Taejon, Korea in October, 1997. Eight highly professed foreign experts whose fields of expertise projected to the area of high level waste disposal were invited to the symposium. This study is composed of four major areas; disposal system design/construction, engineered barrier characterization, geologic environment evaluation and performance assessment and total safety. A technical tree scheme of HLW disposal has been illustrated according to the investigation and an analysis for each technical area. For each detailed technology, research projects, performing organization/method and techniques that are to be secured in the order of priority are proposed, but the suggestions are merely at a superfluous level of propositional idea due to the reduction of the budget in the second year. The detailed programs on HLW disposal are greatly affected by governmental HLW disposal policy and in this study, the primary decisions to be made in each level of HLW disposal enterprise and a rough scheme are proposed. (author). 20 refs., 97 figs., 33 tabs.

  10. Impact of Alkali Source on Vitrification of SRS High Level Waste

    International Nuclear Information System (INIS)

    LAMBERT, D. P.; MILLER, D. H.; PEELER, D. K.; SMITH, M. E.; STONE, M. E.

    2005-01-01

    The Defense Waste Processing Facility (DWPF) Savannah River Site is currently immobilizing high level nuclear waste sludge by vitrification in borosilicate glass. The processing strategy involves blending a large batch of sludge into a feed tank, washing the sludge to reduce the amount of soluble species, then processing the large ''sludge batch'' through the DWPF. Each sludge batch is tested by the Savannah River National Laboratory (SRNL) using simulants and tests with samples of the radioactive waste to ''qualify'' the batch prior to processing in the DWPF. The DWPF pretreats the sludge by first acidifying the sludge with nitric and formic acid. The ratio of nitric to formic acid is adjusted as required to target a final glass composition that is slightly reducing (the target is for ∼20% of the iron to have a valence of two in the glass). The formic acid reduces the mercury in the feed to elemental mercury which is steam stripped from the feed. After a concentration step, the glass former (glass frit) is added as a 50 wt% slurry and the batch is concentrated to approximately 50 wt% solids. The feed slurry is then fed to a joule heated melter maintained at 1150 C. The glass must meet both processing (e.g., viscosity and liquidus temperature) and product performance (e.g., durability) constraints The alkali content of the final waste glass is a critical parameter that affects key glass properties (such as durability) as well as the processing characteristics of the waste sludge during the pretreatment and vitrification processes. Increasing the alkali content of the glass has been shown to improve the production rate of the DWPF, but the total alkali in the final glass is limited by constraints on glass durability and viscosity. Two sources of alkali contribute to the final alkali content of the glass: sodium salts in the waste supernate and sodium and lithium oxides in the glass frit added during pretreatment processes. Sodium salts in the waste supernate can

  11. Hanford tanks initiative plan

    International Nuclear Information System (INIS)

    McKinney, K.E.

    1997-01-01

    Abstract: The Hanford Tanks Initiative (HTI) is a five-year project resulting from the technical and financial partnership of the U.S. Department of Energy's Office of Waste Management (EM-30) and Office of Science and Technology Development (EM-50). The HTI project accelerates activities to gain key technical, cost performance, and regulatory information on two high-level waste tanks. The HTI will provide a basis for design and regulatory decisions affecting the remainder of the Tank Waste Remediation System's tank waste retrieval Program

  12. Development, evaluation, and selection of candidate high-level waste forms

    International Nuclear Information System (INIS)

    Bernadzikowski, T.A.; Allender, J.S.; Gordon, D.E.; Gould, T.H. Jr.

    1982-01-01

    The seven candidate waste forms, evaluated as potential media for the immobilization and gelogic disposal of high-level nuclear wastes were borosilicate glass, SYNROC, tailored ceramic, high-silica glass, FUETAP concrete, coated sol-gel particles, and glass marbles in a lead matrix. The evaluation, completed on August 1, 1981, combined preliminary waste form evaluations conducted at Department of Energy (DOE) defense waste-sites and at independent laboratories, peer review assessments, a product performance evaluation, and a processability analysis. Based on the combined results of these four inputs, two of the seven forms, borosilicate glass and a titanate-based ceramic, SYNROC, were selected as the reference and alternative forms, respectively, for continued development and evaluation in the National HLW Program. The borosilicate glass and ceramic forms were further compared during FY-1982 on the basis of risk assessments, cost comparisons, properties comparisons, and conformance with proposed regulatory and repository criteria. Both the glass and ceramic forms are viable candidates for use at DOE defense HLW sites; they are also candidates for immobilization of commercial reprocessing wastes. This paper describes the waste form screening process, discusses each of the four major inputs considered in the selection of the two forms in 1981, and presents a brief summary of the comparisons of the two forms during 1982 and the selection process to determine the final form for SRP defense HLW

  13. High-level waste glass compendium; what it tells us concerning the durability of borosilicate waste glass

    International Nuclear Information System (INIS)

    Cunnane, J.C.; Allison, J.

    1993-01-01

    Facilities for vitrification of high-level nuclear waste in the United States are scheduled for startup in the next few years. It is, therefore, appropriate to examine the current scientific basis for understanding the corrosion of high-level waste borosilicate glass for the range of service conditions to which the glass products from these facilities may be exposed. To this end, a document has been prepared which compiles worldwide information on borosilicate waste glass corrosion. Based on the content of this document, the acceptability of canistered waste glass for geological disposal is addressed. Waste glass corrosion in a geologic repository may be due to groundwater and/or water vapor contact. The important processes that determine the glass corrosion kinetics under these conditions are discussed based on experimental evidence from laboratory testing. Testing data together with understanding of the long-term corrosion kinetics are used to estimate radionuclide release rates. These rates are discussed in terms of regulatory performance standards

  14. Human factors, system safety, and systems engineering in the transportation of U.S. high-level waste

    International Nuclear Information System (INIS)

    Price, D.L.; Chu, S.C.

    1993-01-01

    The U.S. Nuclear Waste Technical Review Board is an independent agency charged with evaluating the technical and scientific validity of the U.S. Department of Energy's program to manage the disposal of spent fuel and defense high-level waste. The Board has continued to emphasize the importance of using a true system approach in designing the waste management system. The Board has recommended the application of basic design disciplines such as human factors, system safety, and systems engineering. A top-level system study needs to be undertaken that focuses on minimizing handling. The analysis must be well done, in a timely manner, and without the inclusion in the analysis of arbitrary and artificial constraints. (author)

  15. Report on the joint USA-Germany drop test program for a vitrified high level waste cask

    International Nuclear Information System (INIS)

    Golliher, K.G.; Witt, C.R.; Wieser, K.E.

    1993-01-01

    A series of full-scale drop tests was performed on a ductile iron transport cask in a cooperative program between the US Department of Energy (DOE) and Bundesantalt fuer Materialpruefung (BAM) in Germany. The tests, which were performed at BAM's test facility located near Lehre, Germany, were performed on a prototype cask designed for transport of Vitrified High Level Waste (VHLW) canisters. The VHLW cask is a right circular cylinder with a diameter of 1156 mm and a height of 3454 cm, and weighs approximately 24.6 kg including its payload of a single VHLW canister. The drop tests were performed with a non-radioactive, prototype VHLW canister in the cavity. (J.P.N.)

  16. Hanford Waste Vitrification Plant Quality Assurance Program description for defense high-level waste form development and qualification

    International Nuclear Information System (INIS)

    Hand, R.L.

    1992-01-01

    This document describes the quality assurance (QA) program of the Hanford Waste Vitrification Plant (HWVP) Project. The purpose of the QA program is to control project activities in such a manner as to achieve the mission of the HWVP Project in a safe and reliable manner. A major aspect of the HWVP Project QA program is the control of activities that relate to high-level waste (HLW) form development and qualification. This document describes the program and planned actions the Westinghouse Hanford Company (Westinghouse Hanford) will implement to demonstrate and ensure that the HWVP Project meets the US Department of Energy (DOE) and ASME regulations. The actions for meeting the requirements of the Waste Acceptance Preliminary Specifications (WAPS) will be implemented under the HWVP product qualification program with the objective of ensuring that the HWVP and its processes comply with the WAPS established by the federal repository

  17. Melter feed viscosity during conversion to glass: Comparison between low-activity waste and high-level waste feeds

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Tongan [Pacific Northwest National Laboratory, Richland Washington; Chun, Jaehun [Pacific Northwest National Laboratory, Richland Washington; Dixon, Derek R. [Pacific Northwest National Laboratory, Richland Washington; Kim, Dongsang [Pacific Northwest National Laboratory, Richland Washington; Crum, Jarrod V. [Pacific Northwest National Laboratory, Richland Washington; Bonham, Charles C. [Pacific Northwest National Laboratory, Richland Washington; VanderVeer, Bradley J. [Pacific Northwest National Laboratory, Richland Washington; Rodriguez, Carmen P. [Pacific Northwest National Laboratory, Richland Washington; Weese, Brigitte L. [Pacific Northwest National Laboratory, Richland Washington; Schweiger, Michael J. [Pacific Northwest National Laboratory, Richland Washington; Kruger, Albert A. [U.S. Department of Energy, Office of River Protection, Richland Washington; Hrma, Pavel [Pacific Northwest National Laboratory, Richland Washington

    2017-12-07

    During nuclear waste vitrification, a melter feed (generally a slurry-like mixture of a nuclear waste and various glass forming and modifying additives) is charged into the melter where undissolved refractory constituents are suspended together with evolved gas bubbles from complex reactions. Knowledge of flow properties of various reacting melter feeds is necessary to understand their unique feed-to-glass conversion processes occurring within a floating layer of melter feed called a cold cap. The viscosity of two low-activity waste (LAW) melter feeds were studied during heating and correlated with volume fractions of undissolved solid phase and gas phase. In contrast to the high-level waste (HLW) melter feed, the effects of undissolved solid and gas phases play comparable roles and are required to represent the viscosity of LAW melter feeds. This study can help bring physical insights to feed viscosity of reacting melter feeds with different compositions and foaming behavior in nuclear waste vitrification.

  18. Comparison of elastic and inelastic analysis and test results for the defense high level waste shipping cask

    International Nuclear Information System (INIS)

    Zimmer, A.; Koploy, M.A.; Madsen, M.M.

    1991-01-01

    In the early 1980s, the US DOE/Defense Programs (DOE/DP) initiated a project to develop a safe and efficient transportation system for defense high level waste (DHLW). A long-standing objective of the DHLW transportation project is to develop a truck cask that represents the leading edge of cask technology as well as fully complies with all applicable DOE, Nuclear Regulatory Commission, and DOT regulations. General Atomics designed the DHLW Truck Shipping Cask using state-of-the-art analytical techniques verified by model testing performed by Sandia National Labs. (SNL). The analytical techniques include two approaches, inelastic analysis and elastic analysis. This paper will compare the results of the two analytical approaches and with model testing results. The purpose of this work is to provide data to support licensing of the DHLW cask and to support the acceptance by the NRC of inelastic analysis as a tool in packaging design and licensing

  19. Disposal of high-level waste from nuclear power plants in Denmark. Salt dome investigations. v.1

    International Nuclear Information System (INIS)

    1981-01-01

    A summary is presented of a report in five volumes on possible disposal of radioactive waste in Denmark. The investigation was made by the Danish electric utilities ELKRAFT and ELSAM at the request of the Danish Government. The investigation proved it possible to consider two alternative designs for a disposal facility, one based on the deposition of waste in individual, deep holes, the other on placing the waste in mine galleries. A safety analysis was completed with the Mors dome as example. The purpose of the analysis was to prove whether safe disposal of high-level waste in Denmark was feasible. The utilities concluded that the results of the analysis were satisfactory and the report is now being assessed by the authorities. (BP)

  20. Potential role of ABC-assisted repositories in U.S. plutonium and high-level waste disposition

    Energy Technology Data Exchange (ETDEWEB)

    Berwald, D.; Favale, A.; Myers, T. [Grumman Aerospace Corporation, Bethpage, NY (United States)] [and others

    1995-10-01

    This paper characterizes the issues involving deep geologic disposal of LWR spent fuel rods, then presents results of an investigation to quantify the potential role of Accelerator-Based Conversion (ABC) in an integrated national nuclear materials and high level waste disposition strategy. The investigation used the deep geological repository envisioned for Yucca Mt., Nevada as a baseline and considered complementary roles for integrated ABC transmutation systems. The results indicate that although a U.S. geologic waste repository will continue to be required, waste partitioning and accelerator transmutation of plutonium, the minor actinides, and selected long-lived fission products can result in the following substantial benefits: plutonium burndown to near zero levels, a dramatic reduction of the long term hazard associated with geologic repositories, an ability to place several-fold more high level nuclear waste in a single repository, electricity sales to compensate for capital and operating costs.

  1. Progress report on safety research of high-level waste management for the period April 1985 to March 1986

    International Nuclear Information System (INIS)

    Nakamura, Haruto; Tashiro, Shingo

    1986-09-01

    Researches on high level waste management in fiscal year of 1985 is reviewed. Topics are as follows; 1) Glass waste form was examined with emphasis on the leaching mechanisms under various conditions to predict the long-term leach rates. Leaching rate was examined in synthesized groundwater and a leaching model was developed. Cooperation between Japan and Australia on development of SYNROC has started. 2) Heating experiments with a real size simulated canister and migration tests using non-sorbing tracer has been carried out in a near surface granite rock mass. 2D-SEEP, the coupled computer code of heat and groundwater flow, has been developed. 3) Japanese group participated the ESOPE project of OECD/NEA SWG. Small particles of an organic phase was found in sediment as a material to have significant influence on technetium fixation. 4) Alpha radiation stability of vitrified forms under beta and gamma irradiation have newly started in WASTEF. (author)

  2. Progress report on safety research on high-level waste management for the period April 1989 to March 1990

    International Nuclear Information System (INIS)

    Muraoka, Susumu; Senoo, Muneaki; Kobayashi, Yoshii

    1991-02-01

    Research on high-level waste management at the Engineered Barrier Materials Laboratory, Environmental Geochemistry Laboratory and Environmental Radiochemistry Laboratory of the Department of Environmental Safety Research, JAERI in the fiscal year of 1989 are described. The topics are as follows: 1) As for waste forms and engineered barrier material, performance assessment studies on glass and ceramic forms, and corrosion test of carbon steel were continued. 2) In the safety evaluation study for geological disposal, chemical behavior of nuclide in water, nuclide migration and retardation in geosphere were studied. New microspectrometers was developed to analyze the chemical form in rocks. 3) Distribution and migration of uranium in uranium ore were examined as a natural analogue study. (author)

  3. Comparison of SRP high-level waste disposal costs for borosilicate glass and crystalline ceramic waste forms

    International Nuclear Information System (INIS)

    McDonell, W.R.

    1982-04-01

    An evaluation of costs for the immobilization and repository disposal of SRP high-level wastes indicates that the borosilicate glass waste form is less costly than the crystalline ceramic waste form. The wastes were assumed immobilized as glass with 28% waste loading in 10,300 reference 24-in.-diameter canisters or as crystalline ceramic with 65% waste loading in either 3400 24-in.-diameter canisters or 5900 18-in.-diameter canisters. After an interim period of onsite storage, the canisters would be transported to the federal repository for burial. Total costs in undiscounted 1981 dollars of the waste disposal operations, excluding salt processing for which costs are not yet well defined, were about $2500 million for the borosilicate glass form in reference 24-in.-diameter canisters, compared to about $2900 million for the crystalline ceramic form in 24-in.-diameter canisters and about $3100 million for the crystalline ceramic form in 18-in.-diameter canisters. No large differences in salt processing costs for the borosilicate glass and crystalline ceramic forms are expected. Discounting to present values, because of a projected 2-year delay in startup of the DWPF for the crystalline ceramic form, preserved the overall cost advantage of the borosilicate glass form. The waste immobilization operations for the glass form were much less costly than for the crystalline ceramic form. The waste disposal operations, in contrast, were less costly for the crystalline ceramic form, due to fewer canisters requiring disposal; however, this advantage was not sufficient to offset the higher development and processing costs of the crystalline ceramic form. Changes in proposed Nuclear Regulatory Commission regulations to permit lower cost repository packages for defense high-level wastes would decrease the waste disposal costs of the more numerous borosilicate glass forms relative to the crystalline ceramic forms

  4. Shielding properties of protective thin film coatings and blended concrete compositions for high level waste storage packages

    International Nuclear Information System (INIS)

    Fusco, Michael A.; Winfrey, Leigh; Bourham, Mohamed A.

    2016-01-01

    Highlights: • Measured linear attenuation coefficients are the same for bare and coated steels. • Gamma mean free path is much larger than coating thickness; buildup is negligible. • ‘Concrete-6’ reduces exposure rate outside spent fuel cask significantly over ordinary concrete. - Abstract: Various thin film coatings have been proposed to protect stainless steel high level waste (HLW) containers from premature failure due to localized corrosion, hydrogen embrittlement, and mechanical wear. These coatings include TiN, ZrO 2 , MoS 2 , TiO 2 , and Al 2 O 3 , to be deposited either in multiple layers or as a thicker, single-layer composite. Linear attenuation coefficients of these materials have been simulated using MicroShield and measured experimentally for various photon energies. Additionally, spent fuel casks with overpacks made of two different types of concrete were simulated to compare exposure rate at the cask surface. In the energy range that is significant for high level waste storage all coating materials possess very similar attenuation behavior. A specialty concrete, containing magnetite (Fe 3 O 4 ) and lead oxide (PbO), reduces the exposure rate at the outer surface of the overpack by several orders of magnitude. The higher-Z elements not present in ordinary concrete greatly increase attenuation of intermediate-energy gammas (0.4–1.0 MeV). The thin film coatings do not affect the shielding capabilities of the HLW packaging, as their total proposed thickness is nearly three orders of magnitude less than the mean free path (MFP) of the primary photons of interest.

  5. Initial demonstration of the NRC`s capability to conduct a performance assessment for a High-Level Waste Repository

    Energy Technology Data Exchange (ETDEWEB)

    Codell, R.; Eisenberg, N.; Fehringer, D.; Ford, W.; Margulies, T.; McCartin, T.; Park, J.; Randall, J.

    1992-05-01

    In order to better review licensing submittals for a High-Level Waste Repository, the US Nuclear Regulatory Commission staff has expanded and improved its capability to conduct performance assessments. This report documents an initial demonstration of this capability. The demonstration made use of the limited data from Yucca Mountain, Nevada to investigate a small set of scenario classes. Models of release and transport of radionuclides from a repository via the groundwater and direct release pathways provided preliminary estimates of releases to the accessible environment for a 10,000 year simulation time. Latin hypercube sampling of input parameters was used to express results as distributions and to investigate model sensitivities. This methodology demonstration should not be interpreted as an estimate of performance of the proposed repository at Yucca Mountain, Nevada. By expanding and developing the NRC staff capability to conduct such analyses, NRC would be better able to conduct an independent technical review of the US Department of Energy (DOE) licensing submittals for a high-level waste (HLW) repository. These activities were divided initially into Phase 1 and Phase 2 activities. Additional phases may follow as part of a program of iterative performance assessment at the NRC. The NRC staff conducted Phase 1 activities primarily in CY 1989 with minimal participation from NRC contractors. The Phase 2 activities were to involve NRC contractors actively and to provide for the transfer of technology. The Phase 2 activities are scheduled to start in CY 1990, to allow Sandia National Laboratories to complete development and transfer of computer codes and the Center for Nuclear Waste Regulatory Analyses (CNWRA) to be in a position to assist in the acquisition of the codes.

  6. 45-Day safety screening results for tank 241-U-102, push mode cores 143 and 144

    International Nuclear Information System (INIS)

    Steen, F.H.

    1996-01-01

    This document is the 45-day report deliverable for tank 241-U-102 push mode core segments collected between April 16, 1996 and May 6, 1996 and received by the 222-S Laboratory between April 17, 1996 and May 8, 1996. The segments were subsampled and analyzed in accordance, with the Tank 241-U-102 Push Mode Core Sampling and analysis Plan (TSAP) (Hu, 1996) and the Safety Screening Data Quality Objective (DQO) (Dukelow, et al., 1995). The analytical results are included in Table 1. Attachment I is a cross reference to relate the tank farm identification numbers to the 222-S Laboratory LabCore sample numbers. The subsamples generated in the laboratory for analysis are identified in these diagrams with their sources shown. The diagram identifying the hydrostatic head fluid (HHF) blank is also included, Primary safety screening results and the raw data from Differential Scanning Calorimetry (DSC) and thermogravimetric analysis (TGA) analyses are included in this report. Two of the samples submitted for DSC analysis exceeded notification limits as stated in the Safety Screening DQO (Dukelow, et al., 1995). Cyanide analysis was requested on these samples and a Reactive System Screening Tool analysis was requested for the sample exhibiting the highest exothenn in accordance with the TSAP (Hu, 1996). The results for these analyses will be reported in a revision to this document

  7. Commercial high-level-waste management: options and economics. A comparative analysis of the ceramic and glass waste forms

    International Nuclear Information System (INIS)

    McKisson, R.L.; Grantham, L.F.; Guon, J.; Recht, H.L.

    1983-01-01

    Results of an estimate of the waste management costs of the commercial high-level waste from a 3000 metric ton per year reprocessing plant show that the judicious use of the ceramic waste form can save about $2 billion during a 20-year operating campaign relative to the use of the glass waste form. This assumes PWR fuel is processed and the waste is encapsulated in 0.305-m-diam canisters with ultimate emplacement in a BWIP-type horizontal-borehole repository. The estimated total cost (capital and operating) of the management in the ceramic form is $2.0 billion, and that of the glass form is $4.0 billion. Waste loading and waste form density are the driving factors in that the low-waste loading (25%) and relatively low density (3.1 g/cm 3 ) characteristic of the glass form require several times as many canisters to handle a given waste throughput than is needed for the ceramic waste form whose waste loading capability exceeds 60% and whose waste density is nominally 5.2 g/cm 3 ) characteristic of the glass form requires several times as many canisters to handle a given waste throughput than is needed for the ceramic waste form whose waste loading capability exceeds 60% and whose waste density is nominally 5.2 g/cm 3 . The minimum-cost ceramic waste form has a 60 wt. % waste loading of commercial high-level waste and requires 25 years storage before emplacement in basalt with delayed backfill. Because of the process flexibility allowed by the availability of the high-waste loading of the ceramic form, the intermediate-level liquid waste stream can be mixed with the high-level liquid waste stream and economically processed and emplaced. The cost is greater by $0.3 billion than that of the best high-level liquid waste handling process sequence ($2.3 billion vs $2.0 billion), but this difference is less than the cost of the separate disposal of the intermediate-level liquid waste

  8. Hanford tanks initiative - test implementation plan for demonstration of in-tank retrieval technology

    International Nuclear Information System (INIS)

    Schaus, P.S.

    1997-01-01

    This document presents a Systems Engineering approach for performing the series of tests associated with demonstrating in-tank retrieval technologies. The testing ranges from cold testing of individual components at the vendor's facility to the final fully integrated demonstration of the retrieval system's ability to remove hard heel high-level waste from the bottom of a Hanford single-shell tank

  9. Corrosion behaviour of container materials for geological disposal of high-level waste. Joint annual progress report 1983

    International Nuclear Information System (INIS)

    1985-01-01

    Within the framework of the Community R and D programme on management and storage of radioactive waste (shared-cost action), a research activity is aiming at the assessment of corrosion behaviour of potential container materials for geological disposal of vitrified high-level wastes. In this report, the results obtained during the year 1983 are described. Research performed at the Studiecentrum voor Kernenergie/Centre d'Etudes de l'Energie Nucleaire (SCK/CEN) at Mol (B), concerns the corrosion behaviour in clay environments. The behaviour in salt is tested by the Kernforschungszentrum (KfK) at Karlsruhe (D). Corrosion behaviour in granitic environments is being examined by the Commissariat a l'Energie Atomique (CEA) at Fontenay-aux-Roses (F) and the Atomic Energy Research Establishment (AERE) at Harwell (UK); the first is concentrating on corrosion-resistant materials and the latter on corrosion-allowance materials. Finally, the Centre National de la Recherche Scientifique (CNRS) at Vitry (F) is examining the formation and behaviour of passive layers on the metal alloys in the various environments

  10. Development of the high-level waste high-temperature melter feed preparation flowsheet for vitrification process testing

    International Nuclear Information System (INIS)

    Seymour, R.G.

    1995-01-01

    High-level waste (HLW) feed preparation flowsheet development was initiated in fiscal year (FY) 1994 to evaluate alternative flowsheets for preparing melter feed for high-temperature melter (HTM) vitrification testing. Three flowsheets were proposed that might lead to increased processing capacity relative to the Hanford Waste Vitrification Plant (HWVP) and that were flexible enough to use with other HLW melter technologies. This document describes the decision path that led to the selection of flowsheets to be tested in the FY 1994 small-scale HTM tests. Feed preparation flowsheet development for the HLW HTM was based on the feed preparation flowsheet that was developed for the HWVP. This approach allowed the HLW program to build upon the extensive feed preparation flowsheet database developed under the HWVP Project. Primary adjustments to the HWVP flowsheet were to the acid adjustment and glass component additions. Developmental background regarding the individual features of the HLW feed preparation flowsheets is provided. Applicability of the HWVP flowsheet features to the new HLW vitrification mission is discussed. The proposed flowsheets were tested at the laboratory-scale at Pacific Northwest Laboratory. Based on the results of this testing and previously established criteria, a reductant-based flowsheet using glycolic acid and a nitric acid-based flowsheet were selected for the FY 1994 small-scale HTM testing

  11. Experiences from risk communication in the siting of a geological repository for high level waste in Sweden

    International Nuclear Information System (INIS)

    Thegerstroem, C.; Engstroem, S.

    1999-01-01

    SKB is planning in the year 2001 to designate two siting alternatives for further site characterisation. The work in the municipalities of Oesthammar, Nykoeping, Oskarshamn and Tierp is taking place in an atmosphere of constructive discussions. There is a growing feeling in Sweden among broad categories of the public that the nuclear waste exists and should be taken care of by our generation, without many of these people ever getting positive to the use of nuclear energy. While the NIMBY syndrome might still have a good grip on some, there has never been a more constructive debate about the nuclear waste than now, even though there still is a lot of work to do. Siting a geological repository for high level waste puts our democratic system under hard tests. The decision making process is about openness, skills in interacting with the public, respect of people's fears and concerns and at last but not the least independent, competent and visible participation by other stakeholders (politicians locally and nationally, regulatory bodies etc). Good skills in risk communication are important ingredients that might facilitate SKB's task as a developer. Far more important however, is the trust we might get from past and present record of handling the waste and from the way we work and behave in the feasibility studies in the municipalities where SKB is involved

  12. A natural analogue for high-level waste in tuff: Chemical analysis and modeling of the Valles site

    International Nuclear Information System (INIS)

    Stockman, H.W.; Krumhansl, J.L.; Ho, C.K.; Kovach, L.; McConnell, V.S.

    1995-01-01

    The contact between an obsidian flow and a steep-walled tuff canyon was examined as an analogue for a high-level waste repository. The analogue site is located in the Valles Caldera in New Mexico, where a massive obsidian flow filled a paleocanyon in the Battleship Rock Tuff. The obsidian flow provided a heat source, analogous to waste panels or an igneous intrusion in a repository, and caused evaporation and migration of water. The tuff and obsidian samples were analyzed for major and trace elements and mineralogy by INAA, XRF, x-ray diffraction, and scanning electron microscopy and electron microprobe. Samples were also analyzed for D/H and 39 Ar/ 40 Ar isotopic composition. Overall, the effects of the heating event seem to have been slight and limited to the tuff nearest the contact. There is some evidence of devitrification and migration of volatiles in the tuff within 10 m of the contact, but variations in major and trace element chemistry are small and difficult to distinguish from the natural (pre-heating) variability of the rocks

  13. The implications of RCRA [Resource Conservation and Recovery Act] regulation for the disposal of transuranic and high-level waste

    International Nuclear Information System (INIS)

    Sigmon, C.F.; Sharples, F.E.; Smith, E.D.

    1988-01-01

    In May of 1987 the Department of Energy (DOE) published a rule interpreting the definition of ''byproduct'' under the Atomic Energy Act. This byproduct rule clarified the role of the Resource Conservation and Recovery Act (RCRA) in the regulation of DOE's radioactive waste management activities. According to the rule, only the radioactive portion of DOE's mixed radioactive and hazardous waste (mixed waste), including mixed transuranic (TRU) and high-level waste (HLW), is exempt from RCRA under the byproduct exemption. The portion of a waste that is hazardous as defined by RCRA is subject to full regulation under RCRA. Because the radioactive and hazardous portions of m any, if not most, DOE wastes are likely to be inseparable, the rule in effect makes most mixed wastes subject to dual regulation. The potential application of RCRA to facilities such as the Waste Isolation Pilot Plant (WIPP) and the HLW repository creates unique challenges for both the DOE and regulatory authorities. Strategies must be developed to assure compliance with RCRA without either causing excessive administrative burdens or abandoning the goal of minimizing radiation exposure. This paper will explore some of the potential regulatory options for and recent trends in the regulation of TRU and HLW under RCRA

  14. Partitioning of actinide from simulated high level wastes arising from reprocessing of PHWR fuels: counter current extraction studies using CMPO

    International Nuclear Information System (INIS)

    Deshingkar, D.S.; Chitnis, R.R.; Wattal, P.K.; Theyyunni, T.K.; Nair, M.K.T.; Ramanujam, A.; Dhami, P.S.; Gopalakrishnan, V.; Rao, M.K.; Mathur, J.N.; Murali, M.S.; Iyer, R.H.; Badheka, L.P.; Banerji, A.

    1994-01-01

    High level wastes (HLW) arising from reprocessing of pressurised heavy water reactor (PHWR) fuels contain actinides like neptunium, americium and cerium which are not extracted in the Purex process. They also contain small quantities of uranium and plutonium in addition to fission products. Removal of these actinides prior to vitrification of HLW can effectively reduce the active surveillance period of final waste form. Counter current studies using indigenously synthesised octyl (phenyl)-N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO) were taken up as a follow-up of successful runs with simulated sulphate bearing low acid HLW solutions. The simulated HLW arising from reprocessing of PHWR fuel was prepared based on presumed burnup of 6500 MWd/Te of uranium, 3 years cooling period and 800 litres of waste generation per tonne of fuel reprocessed. The alpha activity of the HLW raffinate after extraction with the CMPO-TBP mixture could be brought down to near background level. (author). 13 refs., 2 tabs., 12 figs

  15. Annotated bibliography for the design of waste packages for geologic disposal of spent fuel and high-level waste

    International Nuclear Information System (INIS)

    Wurm, K.J.; Miller, N.E.

    1982-11-01

    This bibliography identifies documents that are pertinent to the design of waste packages for geologic disposal of nuclear waste. The bibliography is divided into fourteen subject categories so that anyone wishing to review the subject of leaching, for example, can turn to the leaching section and review the abstracts of reports which are concerned primarily with leaching. Abstracts are also cross referenced according to secondary subject matter so that one can get a complete list of abstracts for any of the fourteen subject categories. All documents which by their title alone appear to deal with the design of waste packages for the geologic disposal of spent fuel or high-level waste were obtained and reviewed. Only those documents which truly appear to be of interest to a waste package designer were abstracted. The documents not abstracted are listed in a separate section. There was no beginning date for consideration of a document for review. About 1100 documents were reviewed and about 450 documents were abstracted

  16. Direct conversion of surplus fissile materials, spent nuclear fuel, and other materials to high-level-waste glass

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Elam, K.R.

    1995-01-01

    With the end of the cold war the United States, Russia, and other countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. The United States Academy of Sciences (NAS) has recommended that these surplus fissile materials (SFMs) be processed so they are no more accessible than plutonium in spent nuclear fuel (SNF). This spent fuel standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. The NAS recommended investigation of three sets of options for disposition of SFMs while meeting the spent fuel standard: (1) incorporate SFMs with highly radioactive materials and dispose of as waste, (2) partly burn the SFMs in reactors with conversion of the SFMs to SNF for disposal, and (3) dispose of the SFMs in deep boreholes. The US Government is investigating these options for SFM disposition. A new method for the disposition of SFMs is described herein: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptinium, americium, and 233 U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal

  17. Annotated bibliography for the design of waste packages for geologic disposal of spent fuel and high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Wurm, K.J.; Miller, N.E.

    1982-11-01

    This bibliography identifies documents that are pertinent to the design of waste packages for geologic disposal of nuclear waste. The bibliography is divided into fourteen subject categories so that anyone wishing to review the subject of leaching, for example, can turn to the leaching section and review the abstracts of reports which are concerned primarily with leaching. Abstracts are also cross referenced according to secondary subject matter so that one can get a complete list of abstracts for any of the fourteen subject categories. All documents which by their title alone appear to deal with the design of waste packages for the geologic disposal of spent fuel or high-level waste were obtained and reviewed. Only those documents which truly appear to be of interest to a waste package designer were abstracted. The documents not abstracted are listed in a separate section. There was no beginning date for consideration of a document for review. About 1100 documents were reviewed and about 450 documents were abstracted.

  18. FTIR and Mössbauer spectroscopic study of sodium-aluminum-iron phosphate glassy materials for high level waste immobilization

    Science.gov (United States)

    Stefanovsky, S. V.; Stefanovsky, O. I.; Remizov, M. B.; Belanova, E. A.; Kozlov, P. V.; Glazkova, Ya. S.; Sobolev, A. V.; Presniakov, I. A.; Kalmykov, S. N.; Myasoedov, B. F.

    2015-11-01

    Complex sodium-aluminum-iron phosphate glassy materials with various Al2O3 to Fe2O3 ratio containing high level waste (HLW) surrogate were characterized by X-ray diffraction and scanning electron microscopy and studied in details by Fourier transform infrared (FTIR) spectroscopy. The samples with high Al2O3 content and not containing Fe2O3 were predominantly amorphous but subjected to devitrification under annealing. Addition of B2O3 and partial Fe2O3 substitution for Al2O3 in the materials increases their resistance to devitrification whereas further substitution and NiO incorporation significantly increase the tendency to devitrification. FTIR spectra demonstrate changes in the structure of glassy materials caused by both structural variations in the anionic motif and occurrence of crystalline phases in the materials. According to Mössbauer spectroscopy data, iron in the glassy samples is present as octahedrally coordinated Fe3+ ions while in the partly devitrified samples iron is partitioned among vitreous and crystalline phases entering the vitreous phase mainly as Fe3+O6 units and crystalline phases as major Fe3+ and minor Fe2+ ions in a magnetically ordered state and participating in a "fast" electronic exchange.

  19. On the way of advancing technical development related to behavior of bentonite and overpack in high level waste disposal place

    International Nuclear Information System (INIS)

    Toyota, Masatoshi; Ueda, Hiroyoshi.

    1997-01-01

    When the concept on disposal in a high level waste disposal place and the layout including the thickness of bentonite, overpack and so on are investigated, it is necessary to elucidate the dynamic behavior of bentonite and overpack for long period, in particular, the dynamic interaction of bentonite and overpack accompanying the corrosion and expansion of the overpack by its reaction with the groundwater containing little oxygen. At present, the experiment on the dynamic behavior of bentonite is carried out in Power Reactor and Nuclear Fuel Development Corporation, Central Research Institute of Electric Power Industry and others, and the obtained knowledges are reported. The shape of hollow cylinder bentonite specimen is discussed. As for the breaking of overpack accompanying the corrosion and expansion, the equation for evaluating various quantities and the flow chart of the analysis works are shown. The method of judging overpack breaking is explained. The effective thickness of bentonite for safety evaluation is given. If the subjects to be examined hereafter are not concluded by 2000, the alternative of covering overpack with titanium alloy should be investigated in parallel. (K.I.)

  20. Preliminary assessment of the radiological protection aspects of disposal of high-level waste in geologic formations

    International Nuclear Information System (INIS)

    Hill, M.D; Grimwood, P.D.

    1978-01-01

    The purpose of this study is to carry out a preliminary assessment of the potential radiological consequences of disposing of vitrified high-level radioactive waste in geologic formations. The events which could lead to the release of radioactivity from a geologic repository are reviewed and ingress of ground-water is identified as the principal mechanism by which radioactivity may be transported back to the biosphere. A mathematical model of radionuclide migration with ground-water is used to predict possible rates of release of radioactivity into fresh water from a hypothetical repository containing all the high-level waste which may be generated in the UK up to the year 2000. The individual and collective doses which could be received as a result of man's use of contaminated fresh water are evaluated. The numerical results of the study depend very much on the assumptions made and cannot be used to draw any detailed conclusions. The main result is the identification of areas where further studies are required in order to carry out a full evaluation of this disposal option. (author)

  1. Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation

    Energy Technology Data Exchange (ETDEWEB)

    None

    1987-12-01

    The purpose of this report, and the information contained in the associated computerized data bases, is to establish the DOE/OCRWM reference characteristics of the radioactive waste materials that may be accepted by DOE for emplacement in the mined geologic disposal system. This report provides relevant technical data for use by DOE and its supporting contractors and is not intended to be a policy document. This document is backed up by five PC-compatible data bases, written in a user-oriented, menu-driven format, which were developed for this purpose. The data bases are the LWR Assemblies Data Base; the LWR Radiological Data Base; the LWR Quantities Data Base; the LWR NFA Hardware Data Base; and the High-Level Waste Data Base. The above data bases may be ordered using the included form. An introductory information diskette can be found inside the back cover of this report. It provides a brief introduction to each of these five PC data bases. 116 refs., 18 figs., 67 tabs.

  2. Development of the high-level waste high-temperature melter feed preparation flowsheet for vitrification process testing

    Energy Technology Data Exchange (ETDEWEB)

    Seymour, R.G.

    1995-02-17

    High-level waste (HLW) feed preparation flowsheet development was initiated in fiscal year (FY) 1994 to evaluate alternative flowsheets for preparing melter feed for high-temperature melter (HTM) vitrification testing. Three flowsheets were proposed that might lead to increased processing capacity relative to the Hanford Waste Vitrification Plant (HWVP) and that were flexible enough to use with other HLW melter technologies. This document describes the decision path that led to the selection of flowsheets to be tested in the FY 1994 small-scale HTM tests. Feed preparation flowsheet development for the HLW HTM was based on the feed preparation flowsheet that was developed for the HWVP. This approach allowed the HLW program to build upon the extensive feed preparation flowsheet database developed under the HWVP Project. Primary adjustments to the HWVP flowsheet were to the acid adjustment and glass component additions. Developmental background regarding the individual features of the HLW feed preparation flowsheets is provided. Applicability of the HWVP flowsheet features to the new HLW vitrification mission is discussed. The proposed flowsheets were tested at the laboratory-scale at Pacific Northwest Laboratory. Based on the results of this testing and previously established criteria, a reductant-based flowsheet using glycolic acid and a nitric acid-based flowsheet were selected for the FY 1994 small-scale HTM testing.

  3. Prediction of corrosion depth of selected materials for the container of high-level wastes under a repository condition

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Soo; Chun, Kwan Sik; Kang, Chul Hyung; Choi, Jong Won; Han, Kyung Won [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-05-01

    The corrosion depth of selected materials for container of high-level wastes in an underground disposal condition was predicted by analyzing the corrosion behaviors and corrosion rates of copper/copper alloys, carbon steel, titanium/titanium alloys, stainless steel and nickel alloys. Their corrosion rates depend on the amount of oxygen and microbes in bentonite at the bore hole, and local corrosion in addition to general corrosion. However, the effect of radiation and the oxygen dissolved in groundwater would be insignificant. To calculate the corrosion depth, it is assumed that the total amount of oxygen contained in the pore and surface of a bentonite block, and in the gaps among container, rock and bentonite block at a borehole is 300 moles. Assuming that all organic compounds in a bentonite block are presumed as lactate, they would produce 2,100 moles of HS-. The corrosion depths were calculated based on the above assumptions and the wall thickness of copper, carbon steel, titanium, stainless steel and nickel alloys of at least 2.6, 25, 1.3, 5 and 0.3 mm would be required for their corrosion allowances that guarantee their desired service life of 1,000 years. 94 refs., 13 figs., 6 tabs. (Author)

  4. Optical and Microcantilever-Based Sensors for Real-Time In Situ Characterization of High-Level Waste

    International Nuclear Information System (INIS)

    Braun, Gilbert M.; Bryan, Samuel

    2002-01-01

    Fundamental research is being conducted to develop sensors for strontium that can be used in real-time to characterize high-level waste (HLW) process streams. Two fundamentally different approaches are being pursued, which have in common the dependence on highly selective molecular recognition agents. In one approach, an array of chemically selective sensors with sensitive fluorescent probes to signal the presence of the constituent of interest are coupled to fiber optics for remote analytical applications. The second approach employs sensitive microcantilever sensors that have been demonstrated to have unprecedented sensitivity in solution for Cs+ and CrO4 -. Selectivity in microcantilever-based sensors is achieved by modifying the surface of a gold-coated cantilever with a monolayer coating of an alkanethiol derivative of the molecular recognition agent. The approaches are complementary since fiber optic sensors can be deployed in the highly alkaline environment of HLW, bu t a method of immobilizing a fluorescent molecular recognition agents in a polymer film or bead on the surface of the optical fiber has yet to be demonstrated. The microcantilever-based sensors function by converting molecular complexation into surface stress, and they have been demonstrated to have the requisite sensitivity. However, we will investigate method of protecting Si or SiN microcantilever sensors in the highly alkaline environment of HLW while maintaining high selectivity. One objective of this project is to develop Sr(II) molecular recognition agents with rapidly established equilibria needed for real-time analysis, and initial research will focus on calixarene-crown ethers as a platform. Sensors for alkali metal ions, hydroxide, and temperature will be part of the array of sensor elements that will be demonstrated in this program for both the cantilever and fiber optic sensor approaches

  5. Application of the Spanish methodological approach for biosphere assessment to a generic high-level waste disposal site.

    Science.gov (United States)

    Agüero, A; Pinedo, P; Simón, I; Cancio, D; Moraleda, M; Trueba, C; Pérez-Sánchez, D

    2008-09-15

    A methodological approach which includes conceptual developments, methodological aspects and software tools have been developed in the Spanish context, based on the BIOMASS "Reference Biospheres Methodology". The biosphere assessments have to be undertaken with the aim of demonstrating compliance with principles and regulations established to limit the possible radiological impact of radioactive waste disposals on human health and on the environment, and to ensure that future generations will not be exposed to higher radiation levels than those that would be acceptable today. The biosphere in the context of high-level waste disposal is defined as the collection of various radionuclide transfer pathways that may result in releases into the surface environment, transport within and between the biosphere receptors, exposure of humans and biota, and the doses/risks associated with such exposures. The assessments need to take into account the complexity of the biosphere, the nature of the radionuclides released and the long timescales considered. It is also necessary to make assumptions related to the habits and lifestyle of the exposed population, human activities in the long term and possible modifications of the biosphere. A summary on the Spanish methodological approach for biosphere assessment are presented here as well as its application in a Spanish generic case study. A reference scenario has been developed based on current conditions at a site located in Central-West Spain, to indicate the potential impact to the actual population. In addition, environmental change has been considered qualitatively through the use of interaction matrices and transition diagrams. Unit source terms of (36)Cl, (79)Se, (99)Tc, (129)I, (135)Cs, (226)Ra, (231)Pa, (238)U, (237)Np and (239)Pu have been taken. Two exposure groups of infants and adults have been chosen for dose calculations. Results are presented and their robustness is evaluated through the use of uncertainty and

  6. Disposal of high-level waste from nuclear power plants in Denmark. Salt dome investigations. v.2

    International Nuclear Information System (INIS)

    1981-01-01

    The present report deals with the geological investigations performed for determing the feasibility of a repository for high-level waste in a salt dome. It is volume 2 of five volumes that together constitute the final report of the Danish utilities' salt dome investigations. The purpose of the work was to procure a more detailed knowledge of the geology of salt domes in North Jutland on example of Mors. The Mors dome is oval with the two axes of approx. 12.5 km and 8 km respectively. Two deep wells have been drilled into the salt. These wells reach 3400-3500 m below surface. Until a depth of about 3200 m Erslev 2 passes through rock salt of Zechstein 1 which is the oldest evaporite series. However, it could also be interlayed with the slightly younger Zechstein 2. At about 3200 m a marker layer was met with Zechstein 2 salt below. Interpretation of cores and results of downhole electromagnetic and borehole gravimetric measurements show that there is a large area around Erslev 2 which consists of very pure sodium chloride with traces of anhydrite (calcium, sulphate) 1-3%. This area is used for the repository design and safety evaluation. The hydrological conditions existing in the strata above the salt dome (caprock) have been investigated with the help of four hydrogeological wells, placed two each, on two different sites. The cores themselves were taken at various depths in all four holes. With these laboratory methods it has been possible to measure data relevant to hydrology - such as porosity and permeability - as well as geochemistry. (BP)

  7. An assessment of the radiological consequences of disposal of high-level waste in coastal geologic formations

    International Nuclear Information System (INIS)

    Hill, M.D.; Lawson, G.

    1980-11-01

    This study was carried out with the objectives of assessing the potential radiological consequences of entry of circulating ground-water into a high-level waste repository sited on the coast; and comparing the results with those of previous assessments for a repository sited inland. Mathematical models are used to calculate the rate of release of radioactivity into ground-water by leaching, the rates of migration of radionuclides with ground-water from the repository to the sea and the concentrations of radionuclides in sea-water and sea-food as a function of time. Estimates are made of the peak annual collective doses and collective dose commitments which could be received as a result of sea-food consumption. Since there are considerable uncertainties associated with the values of many of the parameters used in the calculations the broad features of the results are more significant than the numerical values of predicted annual doses and collective dose commitments. The results of the assessment show that the rates of migration of radionuclides with ground-water are of primary importance in determining the radiological impact of ground-water ingress. The implications of this result for selection of coastal sites and allocation of research effort are discussed. The comparison of coastal and inland sites suggest that coastal siting may have substantial advantages in terms of the radiological consequences to the public after disposal and that a significant fraction of available research effort should therefore be directed towards investigation of coastal sites. This study has been carried out under contract to the United Kingdom Atomic Energy Authority, Harwell, on behalf of the Commission of the European Communities. (author)

  8. Survey of waste package designs for disposal of high-level waste/spent fuel in selected foreign countries

    International Nuclear Information System (INIS)

    Schneider, K.J.; Lakey, L.T.; Silviera, D.J.

    1989-09-01

    This report presents the results of a survey of the waste package strategies for seven western countries with active nuclear power programs that are pursuing disposal of spent nuclear fuel or high-level wastes in deep geologic rock formations. Information, current as of January 1989, is given on the leading waste package concepts for Belgium, Canada, France, Federal Republic of Germany, Sweden, Switzerland, and the United Kingdom. All but two of the countries surveyed (France and the UK) have developed design concepts for their repositories, but none of the countries has developed its final waste repository or package concept. Waste package concepts are under study in all the countries surveyed, except the UK. Most of the countries have not yet developed a reference concept and are considering several concepts. Most of the information presented in this report is for the current reference or leading concepts. All canisters for the wastes are cylindrical, and are made of metal (stainless steel, mild steel, titanium, or copper). The canister concepts have relatively thin walls, except those for spent fuel in Sweden and Germany. Diagrams are presented for the reference or leading concepts for canisters for the countries surveyed. The expected lifetimes of the conceptual canisters in their respective disposal environment are typically 500 to 1,000 years, with Sweden's copper canister expected to last as long as one million years. Overpack containers that would contain the canisters are being considered in some of the countries. All of the countries surveyed, except one (Germany) are currently planning to utilize a buffer material (typically bentonite) surrounding the disposal package in the repository. Most of the countries surveyed plan to limit the maximum temperature in the buffer material to about 100 degree C. 52 refs., 9 figs

  9. Recommendations for codes and standards to be used for design and fabrication of high level waste canister

    International Nuclear Information System (INIS)

    Bermingham, A.J.; Booker, R.J.; Booth, H.R.; Ruehle, W.G.; Shevekov, S.; Silvester, A.G.; Tagart, S.W.; Thomas, J.A.; West, R.G.

    1978-01-01

    This study identifies codes, standards, and regulatory requirements for developing design criteria for high-level waste (HLW) canisters for commercial operation. It has been determined that the canister should be designed as a pressure vessel without provision for any overpressure protection type devices. It is recommended that the HLW canister be designed and fabricated to the requirements of the ASME Section III Code, Division 1 rules, for Code Class 3 components. Identification of other applicable industry and regulatory guides and standards are provided in this report. Requirements for the Design Specification are found in the ASME Section III Code. It is recommended that design verification be conducted principally with prototype testing which will encompass normal and accident service conditions during all phases of the canister life. Adequacy of existing quality assurance and licensing standards for the canister was investigated. One of the recommendations derived from this study is a requirement that the canister be N stamped. In addition, acceptance standards for the HLW waste should be established and the waste qualified to those standards before the canister is sealed. A preliminary investigation of use of an overpack for the canister has been made, and it is concluded that the use of an overpack, as an integral part of overall canister design, is undesirable, both from a design and economics standpoint. However, use of shipping cask liners and overpack type containers at the Federal repository may make the canister and HLW management safer and more cost effective. There are several possible concepts for canister closure design. These concepts can be adapted to the canister with or without an overpack. A remote seal weld closure is considered to be one of the most suitable closure methods; however, mechanical seals should also be investigated

  10. Transmutation performance analysis on coolant options in a hybrid reactor system design for high level waste incineration

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong-Hee; Siddique, Muhammad Tariq; Kim, Myung Hyun, E-mail: mhkim@khu.ac.kr

    2015-11-15

    Highlights: • Waste transmutation performance was compared and analyzed for seven different coolant options. • Reactions of fission and capture showed big differences depending on coolant options. • Moderation effect significantly affects on energy multiplication, tritium breeding and waste transmutation. • Reduction of radio-toxicities of TRUs showed different trend to coolant choice from performance of waste transmutation. - Abstract: A fusion–fission hybrid reactor (FFHR) is one of the most attractive candidates for high level waste transmutation. The selection of coolant affects the transmutation performance of a FFHR. LiPb coolant, as a conventional coolant for a FFHR, has problems such as reduction in neutron economic and magneto-hydro dynamics (MHD) pressure drop. Therefore, in this work, transmutation performance is evaluated and compared for various coolant options such as LiPb, H{sub 2}O, D{sub 2}O, Na, PbBi, LiF-BeF{sub 2} and NaF-BeF{sub 2} applicable to a hybrid reactor for waste transmutation (Hyb-WT). Design parameters measuring performance of a hybrid reactor were evaluated by MCNPX. They are k{sub eff}, energy multiplication factor, neutron absorption ratio, tritium breeding ratio, waste transmutation ratio, support ratio and radiotoxicity reduction. Compared to LiPb, H{sub 2}O and D{sub 2}O are not suitable for waste transmutation because of neutron moderation effect. Waste transmutation performances with Na and PbBi are similar to each other and not different much from LiPb. Even though molten salt such as LiF-BeF{sub 2} and NaF-BeF{sub 2} is good for avoiding MHD pressure drop problem, waste transmutation performance is dropped compared with LiPb.

  11. Solvent systems containing diglycolamide-functionalized calix[4]arenes in room temperature ionic liquid for metal ion extraction: studies with simulated high-level wastes

    NARCIS (Netherlands)

    Sengupta, A.; Mohapatra, Prasanta K.; Iqbal, M.; Huskens, Jurriaan; Verboom, Willem

    2014-01-01

    The extraction of actinide ions such as Am(III), Pu(IV), Np(IV), U(VI) and Pu(VI) was studied from simulated high-level waste (SHLW) solutions using several diglycolamide-functionalised calix[4]arene (Calix-DGA) ligands dissolved in the room temperature ionic liquid (RTIL) [C8mim][NTf2]. The

  12. New synthesis route for N,N-dimethyl-N,N-dioctyl-2,(2-hexyloxyethyl) malonamide (DMDOHEMA) and its evaluation under simulated high-level waste conditions

    International Nuclear Information System (INIS)

    Patil, Ajay B.; Shinde, Vaishali S.; Pathak, P.N.; Mohapatra, P.K.; Manchanda, V.K.

    2011-01-01

    The primary aim of this study was to develop alternative route of synthesis of DMDOHEMA with better purity and yield. The synthesized product was tested for the extraction behavior of Am, Np, Pu as pure tracers as well as under simulated high-level waste conditions. (author)

  13. Criteria: waste tank isolation and stabilization

    International Nuclear Information System (INIS)

    Metz, W.P.; Ogren, W.E.

    1976-09-01

    The crystallized Hanford high-level wastes stored in single-shell underground tanks consist of sludges and salt cakes covered with supernatural liquor. Purpose of stabilization and isolation is to reduce the releases and losses as a result of a loss of tank integrity. The tanks will be modified so that no inadvertent liquid additions can be made. Criteria for the isolation and stabilization are given and discussed briefly

  14. Criteria: waste tank isolation and stabilization

    Energy Technology Data Exchange (ETDEWEB)

    Metz, W.P.; Ogren, W.E.

    1976-09-01

    The crystallized Hanford high-level wastes stored in single-shell underground tanks consist of sludges and salt cakes covered with supernatural liquor. Purpose of stabilization and isolation is to reduce the releases and losses as a result of a loss of tank integrity. The tanks will be modified so that no inadvertent liquid additions can be made. Criteria for the isolation and stabilization are given and discussed briefly. (DLC)

  15. Task Technical and Quality Assurance Plan for the Characterization of the Tank 40H Samples

    International Nuclear Information System (INIS)

    Wilmarth, W.R.

    2000-01-01

    The High Level Waste Tank Farms store and process high-level liquid wastes from a number of sources including F- and H-Canyons and a recycle stream from the Defense Waste Processing Facility (DWPF). The deposition of sodium aluminosilicate along with sodium diuranate in the 242-16H evaporator system led to the removal of authorization to process High Level Waste containing DWPF recycle. Therefore, High Level Waste Engineering has requested SRTC to perform analysis of the contents of Tank 40H and associated transfers from sludge washing to ensure silicon levels are sufficiently low to allow processing of the supernate through the 3H Evaporator

  16. Laboratory Report on Performance Evaluation of Key Constituents during Pre-Treatment of High Level Waste Direct Feed

    Energy Technology Data Exchange (ETDEWEB)

    Huber, Heinz J.

    2013-06-24

    The analytical capabilities of the 222-S Laboratory are tested against the requirements for an optional start up scenario of the Waste Treatment and Immobilization Plant on the Hanford Site. In this case, washed and in-tank leached sludge would be sent directly to the High Level Melter, bypassing Pretreatment. The sludge samples would need to be analyzed for certain key constituents in terms identifying melter-related issues and adjustment needs. The analyses on original tank waste as well as on washed and leached material were performed using five sludge samples from tanks 241-AY-102, 241-AZ-102, 241-AN-106, 241-AW-105, and 241-SY-102. Additionally, solid phase characterization was applied to determine the changes in mineralogy throughout the pre-treatment steps.

  17. Laboratory Report on Performance Evaluation of Key Constituents during Pre-Treatment of High Level Waste Direct Feed

    International Nuclear Information System (INIS)

    Huber, Heinz J.

    2013-01-01

    The analytical capabilities of the 222-S Laboratory are tested against the requirements for an optional start up scenario of the Waste Treatment and Immobilization Plant on the Hanford Site. In this case, washed and in-tank leached sludge would be sent directly to the High Level Melter, bypassing Pretreatment. The sludge samples would need to be analyzed for certain key constituents in terms identifying melter-related issues and adjustment needs. The analyses on original tank waste as well as on washed and leached material were performed using five sludge samples from tanks 241-AY-102, 241-AZ-102, 241-AN-106, 241-AW-105, and 241-SY-102. Additionally, solid phase characterization was applied to determine the changes in mineralogy throughout the pre-treatment steps

  18. Material selection for Multi-Function Waste Tank Facility tanks

    Energy Technology Data Exchange (ETDEWEB)

    Larrick, A.P.; Blackburn, L.D.; Brehm, W.F.; Carlos, W.C.; Hauptmann, J.P. [Westinghouse Hanford Co., Richland, WA (United States); Danielson, M.J.; Westerman, R.E. [Pacific Northwest Lab., Richland, WA (United States); Divine, J.R. [ChemMet Ltd., West Richland, WA (United States); Foster, G.M. [ICF Kaiser Hanford Co., Richland, WA (United States)

    1995-03-01

    This paper briefly summarizes the history of the materials selection for the US Department of Energy`s high-level waste carbon steel storage tanks. It also provides an evaluation of the materials for the construction of new tanks at the evaluation of the materials for the construction of new tanks at the Multi-Function Waste Tank Facility. The evaluation included a materials matrix that summarized the critical design, fabrication, construction, and corrosion resistance requirements: assessed. each requirement: and cataloged the advantages and disadvantages of each material. This evaluation is based on the mission of the Multi-Function Waste Tank Facility. On the basis of the compositions of the wastes stored in Hanford waste tanks, it is recommended that tanks for the Multi-Function Waste Tank Facility be constructed of ASME SA 515, Grade 70, carbon steel.

  19. Disposal of high-level waste from nuclear power plants in Denmark. Salt dome investigations. v.5

    International Nuclear Information System (INIS)

    1981-01-01

    The present report deals with safety evaluation as part of the investigations regarding a repository for high-level waste in a salt dome. It is volume 5 of five volumes that together constitute the final report on the Danish utilities' salt dome investigations. Two characteristics of the waste are of special importance for the safety evaluation: the encasing of the waste in steel casks with 15 cm thick walls affording protection against corrosion, protecting the surroundings against radiation, and protecting the glass cylinders from mechanical damage resulting from the pressure at the bottom of the disposal hole, and the modest generation of heat in the waste at the time of disposal resulting in a maximum temperature increase in the salt close to the waste of approx. 40 deg. C. These characteristics proved to considerably improve the safety margin with respect to unforeseen circumstances. The character of the salt dome and of the salt in the proposed disposal area offers in itself good protection against contact with the ground water outside the dome. The relatively large depth of 1200 and 2500 m of the salt surface also means that neither dome nor disposal facility will be appreciably influenced by glaciations or earthquakes. The chalk above the proposed disposal area is very tight and to retain radioactive matter effectively even in the precence of high concentrations of NaCL. The safety investigations included a number of natural processes and probable events such as the segregation of crystal water from overlooked salt minerals, faulty sealings of disposal holes, permeable fault zones in the chalk overlying the dome, the risk in connection with human penetration into the dome. These conditions will neither lead to the destruction of the waste casks or to the release of waste from the dome. Leaching of a cavern is the only situation which proved to result in a release of radioactive material to the biosphere, but the resulting doses was found to be small

  20. Role of waste packages in the safety of a high level waste repository in a deep geological formation

    International Nuclear Information System (INIS)

    Bretheau, F.; Lewi, J.

    1990-06-01

    The safety of a radioactive waste disposal facility lays on the three following barriers placed between the radioactive materials and the biosphere: the waste package; the engineered barriers; the geological barrier. The function assigned to each of these barriers in the performance assessment is an option taken by the organization responsible for waste disposal management (ANDRA in France), which must show that: expected performances of each barrier (confinement ability, life-time, etc.) are at least equal to those required to fulfill the assigned function; radiation protection requirements are met in all situations considered as credible, whether they be the normal situation or random event situations. The French waste management strategy is based upon two types of disposal depending on the nature and activity of waste packages: - surface disposal intended for low and medium level wastes having half-lives of about 30 years or less and alpha activity less than 3.7 MBq/kg (0.1 Ci/t), for individual packages and less than 0.37 MBq/kg (0.01 Ci/t) in the average. Deep geological disposal intended for TRU and high level wastes. The conditions of acceptance of packages in a surface disposal site are subject to the two fundamental safety rules no. I.2 and III.2.e. The present paper is only dealing with deep geological disposal. For deep geological repositories, three stages are involved: stage preceding definitive disposal (intermediate storage, transportation, handling, setting up in the disposal cavities); stage subsequent to definitive sealing of the disposal cavities but prior to the end of operation of the repository; stage subsequent to closure of the repository. The role of the geological barrier has been determined as the essential part of long term radioactivity confinement, by a working group, set up by the French safety authorities. Essential technical criteria relating to the choice of a site so defined by this group, are the following: very low permeability

  1. Impact of partitioning and transmutation on high-level waste disposal for the fast breeder reactor fuel cycle

    International Nuclear Information System (INIS)

    Nishihara, Kenji; Oigawa, Hiroyuki; Nakayama, Shinichi; Ono, Kiyoshi; Shiotani, Hiroki

    2010-01-01

    The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large

  2. Tank characterization reference guide

    International Nuclear Information System (INIS)

    De Lorenzo, D.S.; DiCenso, A.T.; Hiller, D.B.; Johnson, K.W.; Rutherford, J.H.; Smith, D.J.; Simpson, B.C.

    1994-09-01

    Characterization of the Hanford Site high-level waste storage tanks supports safety issue resolution; operations and maintenance requirements; and retrieval, pretreatment, vitrification, and disposal technology development. Technical, historical, and programmatic information about the waste tanks is often scattered among many sources, if it is documented at all. This Tank Characterization Reference Guide, therefore, serves as a common location for much of the generic tank information that is otherwise contained in many documents. The report is intended to be an introduction to the issues and history surrounding the generation, storage, and management of the liquid process wastes, and a presentation of the sampling, analysis, and modeling activities that support the current waste characterization. This report should provide a basis upon which those unfamiliar with the Hanford Site tank farms can start their research

  3. Generalized Test Plan for the Vitrification of Simulated High-Level -Waste Calcine in the Idaho National Laboratory‘s Bench -Scale Cold Crucible Induction Melter

    Energy Technology Data Exchange (ETDEWEB)

    Vince Maio

    2011-08-01

    This Preliminary Idaho National Laboratory (INL) Test Plan outlines the chronological steps required to initially evaluate the validity of vitrifying INL surrogate (cold) High-Level-Waste (HLW) solid particulate calcine in INL's Cold Crucible Induction Melter (CCIM). Its documentation and publication satisfies interim milestone WP-413-INL-01 of the DOE-EM (via the Office of River Protection) sponsored work package, WP 4.1.3, entitled 'Improved Vitrification' The primary goal of the proposed CCIM testing is to initiate efforts to identify an efficient and effective back-up and risk adverse technology for treating the actual HLW calcine stored at the INL. The calcine's treatment must be completed by 2035 as dictated by a State of Idaho Consent Order. A final report on this surrogate/calcine test in the CCIM will be issued in May 2012-pending next fiscal year funding In particular the plan provides; (1) distinct test objectives, (2) a description of the purpose and scope of planned university contracted pre-screening tests required to optimize the CCIM glass/surrogate calcine formulation, (3) a listing of necessary CCIM equipment modifications and corresponding work control document changes necessary to feed a solid particulate to the CCIM, (4) a description of the class of calcine that will be represented by the surrogate, and (5) a tentative tabulation of the anticipated CCIM testing conditions, testing parameters, sampling requirements and analytical tests. Key FY -11 milestones associated with this CCIM testing effort are also provided. The CCIM test run is scheduled to be conducted in February of 2012 and will involve testing with a surrogate HLW calcine representative of only 13% of the 4,000 m3 of 'hot' calcine residing in 6 INL Bin Sets. The remaining classes of calcine will have to be eventually tested in the CCIM if an operational scale CCIM is to be a feasible option for the actual INL HLW calcine. This remaining calcine

  4. Perspective on methods to calculate a fee for disposal of defense high-level waste in combined (civilian/defense) repositories

    International Nuclear Information System (INIS)

    1986-12-01

    The Department of Energy intends to send the high-level waste from defense operations to combined civilian/defense repositories for disposal. The federal government must pay a fee to cover its fair share of the cost for the disposal system. This report provides an overview perspective on the defense high-level waste (DHLW) quantities and characteristics and on potential alternatives for calculation and payment of the disposal fee. Information on the DHLW expected from government sites includes the number of waste canisters, radioactivity, thermal decay power, mass of defense reactor fuel, and total electrical energy-equivalents. Ranges in quantities are shown where different operating scenarios are being considered. Several different fee determination methods are described and fees for different quantities of waste are estimated. Information is also included on possible payment alternatives, production and shipping schedules, and credits which could be applied to the fee

  5. Development of a pyro-partitioning process for long-lived radioactive nuclides. Process test for pretreatment of simulated high-level waste containing uranium

    International Nuclear Information System (INIS)

    Kurata, Masateru; Hijikata, Takatoshi; Kinoshita, Kensuke; Inoue, Tadashi

    2000-01-01

    A pyro-partitioning process developed at CRIEPI requires a pre-treatment process to convert high-level liquid waste to chloride. A combination process of denitration and chlorination has been developed for this purpose. Continuous process tests using simulated high-level waste were performed to certify the applicability of the process. Test results indicated a successful material balance sufficient for satisfying pyro-partitioning process criteria. In the present study, process tests using simulated high-level waste containing uranium were also carried out to prove that the pre-treatment process is feasible for uranium. The results indicated that uranium can be converted to chloride appropriate for the pyro-partitioning process. The material balance obtained from the tests is to be used to revise the process flow diagram. (author)

  6. Dissolution of Simulated and Radioactive Savannah River Site High-Level Waste Sludges with Oxalic Acid & Citric Acid Solutions

    International Nuclear Information System (INIS)

    STALLINGS, MARY

    2004-01-01

    sludge solids. We recommend that these results be evaluated further to determine if these solutions contain sufficient neutron poisons. We observed low general corrosion rates in tests in which carbon steel coupons were contacted with solutions of oxalic acid, citric acid and mixtures of oxalic and citric acids. Wall thinning can be minimized by maintaining short contact times with these acid solutions. We recommend additional testing with oxalic and oxalic/citric acid mixtures to measure dissolution performance of sludges that have not been previously dried. This testing should include tests to clearly ascertain the effects of total acid strength and metal complexation on dissolution performance. Further work should also evaluate the downstream impacts of citric acid on the SRS High-Level Waste System (e.g., radiochemical separations in the Salt Waste Processing Facility and addition of organic carbon in the Saltstone and Defense Waste Processing facilities)

  7. Material selection for Multi-Function Waste Tank Facility tanks

    International Nuclear Information System (INIS)

    Carlos, W.C.

    1994-01-01

    This report briefly summarizes the history of the materials selection for the US Department of Energy's high-level waste carbon steel storage tanks. It also provide an evaluation of the materials for the construction of new tanks at the Multi-Function Waste Tank Facility. The evaluation included a materials matrix that summarized the critical design, fabrication, construction, and corrosion resistance requirements; assessed each requirement; and cataloged the advantages and disadvantages of each material. This evaluation is based on the mission of the Multi-Function Waste Tank Facility. On the basis of the compositions of the wastes stored in Hanford waste tanks, it is recommended that tanks for the Multi-Function Waste Tank Facility be constructed of normalized ASME SA 516, Grade 70, carbon steel

  8. Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, R.P. [ed.

    1995-03-01

    This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM.

  9. Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results

    International Nuclear Information System (INIS)

    Rechard, R.P.

    1995-03-01

    This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM

  10. Screening values for Non-Carcinogenic Hanford Waste Tank Vapor Chemicals that Lack Established Occupational Exposure Limits

    Energy Technology Data Exchange (ETDEWEB)

    Poet, Torka S.; Mast, Terryl J.; Huckaby, James L.

    2006-02-06

    Over 1,500 different volatile chemicals have been reported in the headspaces of tanks used to store high-level radioactive waste at the U.S. Department of Energy's Hanford Site. Concern about potential exposure of tank farm workers to these chemicals has prompted efforts to evaluate their toxicity, identify chemicals that pose the greatest risk, and incorporate that information into the tank farms industrial hygiene worker protection program. Established occupation exposure limits for individual chemicals and petroleum hydrocarbon mixtures have been used elsewhere to evaluate about 900 of the chemicals. In this report headspace concentration screening values were established for the remaining 600 chemicals using available industrial hygiene and toxicological data. Screening values were intended to be more than an order of magnitude below concentrations that may cause adverse health effects in workers, assuming a 40-hour/week occupational exposure. Screening values were compared to the maximum reported headspace concentrations.

  11. DOWNSTREAM IMPACTS OF SLUDGE MASS REDUCTION VIA ALUMINUM DISSOLUTION ON DWPF PROCESSING OF SAVANNAH RIVER SITE HIGH LEVEL WASTE - 9382

    Energy Technology Data Exchange (ETDEWEB)

    Pareizs, J; Cj Bannochie, C; Michael Hay, M; Daniel McCabe, D

    2009-01-14

    The SRS sludge that was to become a major fraction of Sludge Batch 5 (SB5) for the Defense Waste Processing Facility (DWPF) contained a large fraction of H-Modified PUREX (HM) sludge, containing a large fraction of aluminum compounds that could adversely impact the processing and increase the vitrified waste volume. It is beneficial to reduce the non-radioactive fraction of the sludge to minimize the number of glass waste canisters that must be sent to a Federal Repository. Removal of aluminum compounds, such as boehmite and gibbsite, from sludge can be performed with the addition of NaOH solution and heating the sludge for several days. Preparation of SB5 involved adding sodium hydroxide directly to the waste tank and heating the contents to a moderate temperature through slurry pump operation to remove a fraction of this aluminum. The Savannah River National Laboratory (SRNL) was tasked with demonstrating this process on actual tank waste sludge in our Shielded Cells Facility. This paper evaluates some of the impacts of aluminum dissolution on sludge washing and DWPF processing by comparing sludge processing with and without aluminum dissolution. It was necessary to demonstrate these steps to ensure that the aluminum removal process would not adversely impact the chemical and physical properties of the sludge which could result in slower processing or process upsets in the DWPF.

  12. Site-specific evaluation of safety issues for high-level waste disposal in crystalline rocks. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Jobmann, M. (ed.) [DBE Technology GmbH, Peine (Germany)

    2016-03-31

    In the past, German research and development (R and D) activities regarding the disposal of radioactive waste, including spent nuclear fuel, focused mainly on domal rock salt because rock salt was the preferred host rock formation. In addition, generic R and D work regarding alternative host rocks (crystalline rocks and claystones) had been performed as well for a long time but with lower intensity. Around the year 2000, as a consequence of the moratorium on the Gorleben site, the Federal Government decided to have argillaceous rocks and crystalline rocks investigated in more detail. As Germany does not have any underground research and host rock characterization facilities, international cooperation received a high priority in the German R and D programme for high-level waste (HLW) disposal in order to increase the knowledge regarding alternative host rocks. Major cornerstones of the cooperation are joint projects and experiments conducted especially in underground research laboratories (URL) in crystalline rocks at the Grimsel Test Site (Switzerland) and the Hard Rock Laboratory (HRL) Aespoe(Sweden) and in argillaceous rocks at the URL Mont Terri (Switzerland) and Bure (France). In 2001, the topic of radioactive waste disposal was integrated into the agreement between the former Russian Ministry of Atomic Energy (Minatom, now Rosatom) and the German Ministry of Labor (BMWA), now Ministry of Economic Affairs and Energy (BMWi), on cooperation regarding R and D on the peaceful utilization of nuclear power (agreement on ''Wirtschaftlich-Technische Zusammenarbeit'' WTZ). The intention was to have a new and interesting opportunity for international R and D cooperation regarding HLW disposal in crystalline rocks and the unique possibility to perform site-specific work, to test the safety demonstration tools available, and to expand the knowledge to all aspects specific to these host rocks. Another motivation for joining this cooperation was the

  13. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM 2009

    Energy Technology Data Exchange (ETDEWEB)

    West, B.; Waltz, R.

    2010-06-21

    Aqueous radioactive wastes from Savannah River Site (SRS) separations and vitrification processes are contained in large underground carbon steel tanks. Inspections made during 2009 to evaluate these vessels and other waste handling facilities along with evaluations based on data from previous inspections are the subject of this report. The 2009 inspection program revealed that the structural integrity and waste confinement capability of the Savannah River Site waste tanks were maintained. All inspections scheduled per LWO-LWE-2008-00423, HLW Tank Farm Inspection Plan for 2009, were completed. All Ultrasonic measurements (UT) performed in 2009 met the requirements of C-ESG-00006, In-Service Inspection Program for High Level Waste Tanks, Rev. 1, and WSRC-TR-2002-00061, Rev.4. UT inspections were performed on Tank 29 and the findings are documented in SRNL-STI-2009-00559, Tank Inspection NDE Results for Fiscal Year 2009, Waste Tank 29. Post chemical cleaning UT measurements were made in Tank 6 and the results are documented in SRNL-STI-2009-00560, Tank Inspection NDE Results Tank 6, Including Summary of Waste Removal Support Activities in Tanks 5 and 6. A total of 6669 photographs were made and 1276 visual and video inspections were performed during 2009. Twenty-Two new leaksites were identified in 2009. The locations of these leaksites are documented in C-ESR-G-00003, SRS High Level Waste Tank Leaksite Information, Rev.4. Fifteen leaksites at Tank 5 were documented during tank wall/annulus cleaning activities. Five leaksites at Tank 6 were documented during tank wall/annulus cleaning activities. Two new leaksites were identified at Tank 19 during waste removal activities. Previously documented leaksites were reactivated at Tanks 5 and 12 during waste removal activities. Also, a very small amount of additional leakage from a previously identified leaksite at Tank 14 was observed.

  14. Evaluating Feed Delivery Performance in Scaled Double-Shell Tanks

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kearn P.; Thien, Michael G.

    2013-11-07

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capability using simulated Hanford High-Level Waste (HLW) formulations. This work represents one of the remaining technical issues with the high-level waste treatment mission at Hanford. The TOCs' ability to adequately mix and sample high-level waste feed to meet the WTP WAC Data Quality Objectives must be demonstrated. The tank mixing and feed delivery must support both TOC and WTP operations. The tank mixing method must be able to remove settled solids from the tank and provide consistent feed to the WTP to facilitate waste treatment operations. Two geometrically scaled tanks were used with a broad spectrum of tank waste simulants to demonstrate that mixing using two rotating mixer jet pumps yields consistent slurry compositions as the tank is emptied in a series of sequential batch transfers. Testing showed that the concentration of slow settling solids in each transfer batch was consistent over a wide range of tank operating conditions. Although testing demonstrated that the concentration of fast settling solids decreased by up to 25% as the tank was emptied, batch-to-batch consistency improved as mixer jet nozzle velocity in the scaled tanks increased.

  15. Evaluating Feed Delivery Performance in Scaled Double-Shell Tanks

    International Nuclear Information System (INIS)

    Lee, Kearn P.; Thien, Michael G.

    2013-01-01

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capability using simulated Hanford High-Level Waste (HLW) formulations. This work represents one of the remaining technical issues with the high-level waste treatment mission at Hanford. The TOCs' ability to adequately mix and sample high-level waste feed to meet the WTP WAC Data Quality Objectives must be demonstrated. The tank mixing and feed delivery must support both TOC and WTP operations. The tank mixing method must be able to remove settled solids from the tank and provide consistent feed to the WTP to facilitate waste treatment operations. Two geometrically scaled tanks were used with a broad spectrum of tank waste simulants to demonstrate that mixing using two rotating mixer jet pumps yields consistent slurry compositions as the tank is emptied in a series of sequential batch transfers. Testing showed that the concentration of slow settling solids in each transfer batch was consistent over a wide range of tank operating conditions. Although testing demonstrated that the concentration of fast settling solids decreased by up to 25% as the tank was emptied, batch-to-batch consistency improved as mixer jet nozzle velocity in the scaled tanks increased

  16. Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories

    Energy Technology Data Exchange (ETDEWEB)

    1983-06-01

    The conceptual waste package interim product specifications and data requirements presented are applicable specifically to the normal borosilicate glass product of the Defense Waste Processing Facility (DWPF). They provide preliminary numerical values for the defense high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available.

  17. A conceptual frame for the development of models for the prediction of the local effects of a high level waste repository

    International Nuclear Information System (INIS)

    Palacios, Elias; Ferreri, J.C.

    1985-01-01

    Safety analyses assocciated with the elimination of radioactive wastes in rock masses assume, in all cases, the existance of wastes which will corrode the waste canisters producing the liberation of radionuclides in the rocky and their ultimate migration towards the biosphere. A conceptual discussion is presented which allows the specification to be met by the models for the prediction of the effects of the emplacement of a high level waste repository located at a depth of 500 m in a granitic rock. Furthermore, the radionuclides giving the largest contribution to the radiological impact are identified. (Author) [es

  18. Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories

    International Nuclear Information System (INIS)

    1983-06-01

    The conceptual waste package interim product specifications and data requirements presented are applicable specifically to the normal borosilicate glass product of the Defense Waste Processing Facility (DWPF). They provide preliminary numerical values for the defense high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available

  19. Accident analysis for high-level waste management alternatives in the US Department of Energy Environmental Restoration and Waste Management Programmatic Environmental Impact Statement

    International Nuclear Information System (INIS)

    Folga, S.; Mueller, C.; Roglans-Ribas, J.

    1994-01-01

    A comparative generic accident analysis was performed for the programmatic alternatives for high-level waste (HLW) management in the US Department of Energy Environmental Restoration and Waste Management Programmatic Environmental Impact Statement (EM PEIS). The key facilities and operations of the five major HLW management phases were considered: current storage, retrieval, pretreatment, treatment, and interim canister storage. A spectrum of accidents covering the risk-dominant accidents was analyzed. Preliminary results are presented for HLW management at the Hanford site. A comparison of these results with those previously advanced shows fair agreement

  20. A review and analysis of European industrial experience in handling LWR [light water reactor] spent fuel and vitrified high-level waste

    International Nuclear Information System (INIS)

    Blomeke, J.O.

    1988-06-01

    The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performances of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States. 79 refs., 71 figs., 10 tabs

  1. Preconceptual design study for solidifying high-level waste: Appendices A, B and C West Valley Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    Hill, O.F. (comp.)

    1981-04-01

    This report presents a preconceptual design study for processing radioactive high-level liquid waste presently stored in underground tanks at Western New York Nuclear Service Center (WNYNSC) near West Valley, New York, and for incorporating the radionculides in that waste into a solid. The high-level liquid waste accumulated from the operation of a chemical reprocessing plant by the Nuclear Fuel Services, Inc. from 1966 to 1972. The high-level liquid waste consists of approximately 560,000 gallons of alkaline waste from Purex process operations and 12,000 gallons of acidic (nitric acid) waste from one campaign of processing thoria fuels by a modified Thorex process (during this campaign thorium was left in the waste). The alkaline waste contains approximately 30 million curies and the acidic waste contains approximately 2.5 million curies. The reference process described in this report is concerned only with chemically processing the high-level liquid waste to remove radionuclides from the alkaline supernate and converting the radionuclide-containing nonsalt components in the waste into a borosilicate glass.

  2. Interim criteria for Organic Watch List tanks at the Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    Babad, S.; Turner, D.A.

    1993-09-01

    This document establishes interim criteria for identifying single-shell radioactive waste storage tanks at the Hanford Site that contain organic chemicals mixed with nitrate/nitrite salts in potentially hazardous concentrations. These tanks are designated as ``organic Watch List tanks.`` Watch List tanks are radioactive waste storage tanks that have the potential for release of high-level waste as a result of uncontrolled increases in temperature or pressure. Organic Watch List tanks are those Watch List tanks that contain relatively high concentrations of organic chemicals. Because of the potential for release of high-level waste resulting from uncontrolled increases in temperature or pressure, the organic Watch List tanks (collectively) constitute a Hanford Site radioactive waste storage tank ``safety issue.``

  3. Interim criteria for Organic Watch List tanks at the Hanford Site

    International Nuclear Information System (INIS)

    Babad, S.; Turner, D.A.

    1993-09-01

    This document establishes interim criteria for identifying single-shell radioactive waste storage tanks at the Hanford Site that contain organic chemicals mixed with nitrate/nitrite salts in potentially hazardous concentrations. These tanks are designated as ''organic Watch List tanks.'' Watch List tanks are radioactive waste storage tanks that have the potential for release of high-level waste as a result of uncontrolled increases in temperature or pressure. Organic Watch List tanks are those Watch List tanks that contain relatively high concentrations of organic chemicals. Because of the potential for release of high-level waste resulting from uncontrolled increases in temperature or pressure, the organic Watch List tanks (collectively) constitute a Hanford Site radioactive waste storage tank ''safety issue.''

  4. Tank closure reducing grout

    Energy Technology Data Exchange (ETDEWEB)

    Caldwell, T.B.

    1997-04-18

    A reducing grout has been developed for closing high level waste tanks at the Savannah River Site in Aiken, South Carolina. The grout has a low redox potential, which minimizes the mobility of Sr{sup 90}, the radionuclide with the highest dose potential after closure. The grout also has a high pH which reduces the solubility of the plutonium isotopes. The grout has a high compressive strength and low permeability, which enhances its ability to limit the migration of contaminants after closure. The grout was designed and tested by Construction Technology Laboratories, Inc. Placement methods were developed by the Savannah River Site personnel.

  5. Tank closure reducing grout

    International Nuclear Information System (INIS)

    Caldwell, T.B.

    1997-01-01

    A reducing grout has been developed for closing high level waste tanks at the Savannah River Site in Aiken, South Carolina. The grout has a low redox potential, which minimizes the mobility of Sr 90 , the radionuclide with the highest dose potential after closure. The grout also has a high pH which reduces the solubility of the plutonium isotopes. The grout has a high compressive strength and low permeability, which enhances its ability to limit the migration of contaminants after closure. The grout was designed and tested by Construction Technology Laboratories, Inc. Placement methods were developed by the Savannah River Site personnel

  6. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM - 2011

    Energy Technology Data Exchange (ETDEWEB)

    West, B.; Waltz, R.

    2012-06-21

    Aqueous radioactive wastes from Savannah River Site (SRS) separations and vitrification processes are contained in large underground carbon steel tanks. Inspections made during 2011 to evaluate these vessels and other waste handling facilities along with evaluations based on data from previous inspections are the subject of this report. The 2011 inspection program revealed that the structural integrity and waste confinement capability of the Savannah River Site waste tanks were maintained. All inspections scheduled per SRR-LWE-2011-00026, HLW Tank Farm Inspection Plan for 2011, were completed. Ultrasonic measurements (UT) performed in 2011 met the requirements of C-ESR-G-00006, In-Service Inspection Program for High Level Waste Tanks, Rev. 3, and WSRC-TR-2002-00061, Rev.6. UT inspections were performed on Tanks 25, 26 and 34 and the findings are documented in SRNL-STI-2011-00495, Tank Inspection NDE Results for Fiscal Year 2011, Waste Tanks 25, 26, 34 and 41. A total of 5813 photographs were made and 835 visual and video inspections were performed during 2011. A potential leaksite was discovered at Tank 4 during routine annual inspections performed in 2011. The new crack, which is above the allowable fill level, resulted in no release to the environment or tank annulus. The location of the crack is documented in C-ESR-G-00003, SRS High Level Waste Tank Leaksite Information, Rev.6.

  7. Hanford site waste tank characterization

    International Nuclear Information System (INIS)

    De Lorenzo, D.S.; Simpson, B.C.

    1994-08-01

    This paper describes the on-going work in the characterization of the Hanford-Site high-level waste tanks. The waste in these tanks was produced as part of the nuclear weapons materials processing mission that occupied the Hanford Site for the first 40 years of its existence. Detailed and defensible characterization of the tank wastes is required to guide retrieval, pretreatment, and disposal technology development, to address waste stability and reactivity concerns, and to satisfy the compliance criteria for the various regulatory agencies overseeing activities at the Hanford Site. The resulting Tank Characterization Reports fulfill these needs, as well as satisfy the tank waste characterization milestones in the Hanford Federal Facility Agreement and Consent Order

  8. Hanford site waste tank characterization

    Energy Technology Data Exchange (ETDEWEB)

    De Lorenzo, D.S.; Simpson, B.C.

    1994-08-01

    This paper describes the on-going work in the characterization of the Hanford-Site high-level waste tanks. The waste in these tanks was produced as part of the nuclear weapons materials processing mission that occupied the Hanford Site for the first 40 years of its existence. Detailed and defensible characterization of the tank wastes is required to guide retrieval, pretreatment, and disposal technology development, to address waste stability and reactivity concerns, and to satisfy the compliance criteria for the various regulatory agencies overseeing activities at the Hanford Site. The resulting Tank Characterization Reports fulfill these needs, as well as satisfy the tank waste characterization milestones in the Hanford Federal Facility Agreement and Consent Order.

  9. Conceptual waste package interim product specifications and data requirements for disposal of glass commercial high-level waste forms in salt geologic repositories

    International Nuclear Information System (INIS)

    1983-10-01

    The conceptual waste package interim product specifications and data requirements presented are applicable to the reference glass composition described in PNL-3838 and carbon steel canister described in ONWI-438. They provide preliminary numerical values for the commercial high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses and regulatory requirements become available. 13 references, 1 figure

  10. Final Report - High Level Waste Vitrification System Improvements, VSL-07R1010-1, Rev 0, dated 04/16/07

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Gan, H.; Pegg, I. L.; Gong, W.; Champman, C. C.; Joseph, I.; Matlack, K. S.

    2013-11-13

    This report describes work conducted to support the development and testing of new glass formulations that extend beyond those that have been previously investigated for the Hanford Waste Treatment and Immobilization Plant (WTP). The principal objective was to investigate maximization of the incorporation of several waste components that are expected to limit waste loading and, consequently, high level waste (HLW) processing rates and canister count. The work was performed with four waste compositions specified by the Office of River Protection (ORP); these wastes contain high concentrations of bismuth, chromium, aluminum, and aluminum plus sodium. The tests were designed to identify glass formulations that maximize waste loading while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM100) tests designed to examine the effects of enhanced glass formulations, increased glass processing temperature, increased crystallinity, and feed solids content on waste processing rate and product quality.

  11. Spent fuel and high level waste: Chemical durability and performance under simulated repository conditions. Results of a coordinated research project 1998-2004

    International Nuclear Information System (INIS)

    2007-10-01

    This publication contains the results of an IAEA Coordinated Research Project (CRP). It provides a basis for understanding the potential interactions of waste form and repository environment, which is necessary for the development of the design and safety case for deep disposal. Types of high level waste matrices investigated include spent fuel, glasses and ceramics. Of particular interest are the experimental results pertaining to ceramic forms such as SYNROC. This publication also outlines important areas for future work, namely, standardized, collaborative experimental protocols for package-release studies, structured development and calibration of predictive models linking the performance of packaged waste and the repository environment, and studies of the long term behaviour of the wastes, including active waste samples

  12. The relationship between the Resource Conservation and Recovery Act and the storage and disposal of spent nuclear fuel and high-level waste

    International Nuclear Information System (INIS)

    Gertz, C.P.; Cloke, P.L.

    1993-01-01

    This paper addresses the potential applicability of the requirements of the Resource Conservation and Recovery Act (RCRA) to the disposal of spent commercial nuclear fuel and of high-level (vitrified) radioactive waste. The Atomic Energy Act of 1954, as amended, and the associated regulations issued by the US NRC provides many requirements that apply to these waste forms and largely, if not entirely, pre-empts the applicability of RCRA. The RCRA would apply only to the non-radioactive components of these wastes, and then only in respect to hazardous components. In view of these restrictions it becomes important to evaluate whether any components of spent fuel or high-level waste are toxic, as defined by the RCRA regulations. Present indications are that they are not and, hence, the US DOE is proceeding on the basis that these wastes and others that may be generated in the future are non-hazardous in respect to RCRA definitions

  13. Evaluation of a gas chromatograph with a novel surface acoustic wave detector (SAW GC) for screening of volatile organic compounds in Hanford waste tank samples

    International Nuclear Information System (INIS)

    Lockrem, L.L.

    1998-01-01

    A novel instrument, a gas chromatograph with a Surface Acoustic Wave Detector (SAW GC), was evaluated for the screening of organic compounds in Hanford tank headspace vapors. Calibration data were developed for the most common organic compounds, and the accuracy and precision were measured with a certified standard. The instrument was tested with headspace samples collected from seven Hanford waste tanks

  14. Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, R.P. [ed.

    1993-12-01

    This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste, as mandated by the Nuclear Waste Policy Act of 1982. The waste forms comprised about 700 metric tons of initial heavy metal (or equivalent units) stored at the INEL: graphite spent fuel, experimental low enriched and highly enriched spent fuel, and high-level waste generated during reprocessing of some spent fuel. Five different waste treatment options were studied; in the analysis, the options and resulting waste forms were analyzed separately and in combination as five waste disposal groups. When the waste forms were studied in combination, the repository was assumed to also contain vitrified high-level waste from three DOE sites for a common basis of comparison and to simulate the impact of the INEL waste forms on a moderate-sized repository, The performance of the waste form was assessed within the context of a whole disposal system, using the U.S. Environmental Protection Agency`s Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 40 CFR 191, promulgated in 1985. Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

  15. Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Rechard, R.P.

    1993-12-01

    This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste, as mandated by the Nuclear Waste Policy Act of 1982. The waste forms comprised about 700 metric tons of initial heavy metal (or equivalent units) stored at the INEL: graphite spent fuel, experimental low enriched and highly enriched spent fuel, and high-level waste generated during reprocessing of some spent fuel. Five different waste treatment options were studied; in the analysis, the options and resulting waste forms were analyzed separately and in combination as five waste disposal groups. When the waste forms were studied in combination, the repository was assumed to also contain vitrified high-level waste from three DOE sites for a common basis of comparison and to simulate the impact of the INEL waste forms on a moderate-sized repository, The performance of the waste form was assessed within the context of a whole disposal system, using the U.S. Environmental Protection Agency's Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 40 CFR 191, promulgated in 1985. Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories

  16. Corrosion control of carbon steel radioactive-liquid storage tanks

    International Nuclear Information System (INIS)

    Chang, Ji Young.

    1997-05-01

    As the West Valley Demonstration Project (WVDP) continues vitrification operation and begins decontamination activities, it is vital to continue to maintain the integrity of the high-level waste tanks and prevent further corrosion that may disrupt the operation. This report describes the current operational status and some corrosion concerns with corresponding control measure recommendations. 14 refs., 5 figs., 6 tabs

  17. Tank vapor mitigation requirements for Hanford Tank Farms

    Energy Technology Data Exchange (ETDEWEB)

    Rakestraw, L.D.

    1994-11-15

    Westinghouse Hanford Company has contracted Los Alamos Technical Associates to listing of vapors and aerosols that are or may be emitted from the High Level Waste (HLW) tanks at Hanford. Mitigation requirements under Federal and State law, as well as DOE Orders, are included in the listing. The lists will be used to support permitting activities relative to tank farm ventilation system up-grades. This task is designated Task 108 under MJB-SWV-312057 and is an extension of efforts begun under Task 53 of Purchase Order MPB-SVV-03291 5 for Mechanical Engineering Support. The results of that task, which covered only thirty-nine tanks, are repeated here to provide a single source document for vapor mitigation requirements for all 177 HLW tanks.

  18. Management of high level waste

    International Nuclear Information System (INIS)

    Culler, F.L. Jr.; Platt, A.M.; Blomeke, J.O.

    1977-01-01

    One of the central problems in nuclear energy is the disposal of wastes. This problem is international in scope; its solution will require intense international cooperation and exchange of data and test results. By the year 2000, worldwide use of nuclear power is estimated to reach between 2000 and 2500 gigawatts (GWe) of electrical generating capacity in countries excluding the USSR and China. Nuclear power stations and supporting fuel cycle facilities will exist on each continent. Safe and internationally accepted methods for the management and isolation of fission products and actinide wastes from these facilities must be developed and demonstrated. Many sites, and hopefully several different geologic formations, will have to be tested and approved for storage of these wastes until they decay to safe levels. This paper estimates the quantity of the various radioactive wastes that will be produced worldwide. The methods for treatment, concentration, fixation, and final storable forms are presented, with particular attention to both the recycle and the disposal of actinides. The transportation of wastes across international boundaries will be required. Estimates of the magnitude of and the complexity of waste movement from points of generation to long-term isolation centers, and the current state of technology available, are given. Ultimately, each type of waste will require a long-term, safe, and publicly accepted repository. This paper concludes with a review of the current programs proposed for ultimate waste isolation, with particular emphasis on storage in geologic formations

  19. An analysis of the factors affecting the hydraulic conductivity and swelling pressure of Kyungju ca-bentonite for use as a clay-based sealing material for a high level waste repository

    International Nuclear Information System (INIS)

    Cho, Won Jin; Lee, Jae Owen; Kwon, Sang Ki

    2012-01-01

    The buffer and backfill are important components of the engineered barrier system in a high-level waste repository, which should be constructed in a hard rock formation at a depth of several hundred meters below the ground surface. The primary function of the buffer and backfill is to seal the underground excavation as a preferred flow path for radionuclide migration from the deposited high-level waste. This study investigates the hydraulic conductivity and swelling pressure of Kyungju Ca-bentonite, which is the candidate material for the buffer and backfill in the Korean reference high-level waste disposal system. The factors that influence the hydraulic conductivity and swelling pressure of the buffer and backfill are analyzed. The factors considered are the dry density, the temperature, the sand content, the salinity and the organic carbon content. The possibility of deterioration in the sealing performance of the buffer and backfill is also assessed.

  20. A first survey of disruption scenarios for a high-level-waste repository at Yucca Mountain, Nevada: Nevada Nuclear Waste Storage Investigations Project

    International Nuclear Information System (INIS)

    Ross, B.

    1987-12-01

    A high-level-waste repository located in unsaturated welded tuff at Yucca Mountain, Nevada, would rely on six different, although not entirely independent, barriers to prevent escape of radioactivity. These barriers are the waste canister, fuel cladding, dissolution of the spent fuel itself, and movement of released contaminants in the unsaturated Topopah Spring welded tuff unit, the unsaturated Calico Hills nonwelded tuff unit, and the saturated tuff aquifer. Fifty-eight processes and events that might affect such a repository were examined. Eighty-four different sequences were identified by which these processes and events could lead to failure of one or more barriers. Sequences that had similar consequences were grouped into 17 categories: direct release, repository flooding, colloid formation, increased water flux through the repository, accelerated fracture flow, water diverted toward the waste package, accelerated dissolution mechanisms, accelerated cladding corrosion mechanisms, accelerated canister corrosion mechanisms, canister breakage, fracture flow in the Topopah Spring welded unit without increased moisture flux, reduced sorption in the Topopah Spring welded unit, water table rise above the Calico Hills nonwelded unit, fracture flow in the Calico Hills nonwelded unit, new discharge points, and faster flow in the saturated zone. 52 refs., 6 figs., 1 tab

  1. Analysis of the effect of variations in parameter values on the predicted radiological consequences of geologic disposal of high-level waste

    International Nuclear Information System (INIS)

    Hill, M.D.

    1979-06-01

    A preliminary assessment of the radiological consequences of geologic disposal of high-level waste (Hill and Grimwood. NRPB-R69 (1978)) identified several areas where further research is required before this disposal option can be fully evaluated. This report is an analysis of the sensitivity of the results of the preliminary assessment to the assumptions made and the values of the parameters used. The parameters considered include the leach rate of the waste, the ground-water velocity, the length of the flow path from the repository to a source of drinking water and the sorption constants of the principle radionuclides. The results obtained by varying these parameters are used to examine the effects of assumptions such as the time at which leaching of the waste begins. The sensitivity analysis shows the relative importance of the waste canisters, the waste form and the geologic barrier to radionuclide migration in determining potential doses. These results are used to identify research priorities, establish preliminary design criteria and indicate developments needed in the mathematical modelling of the movement of radionuclides from a repository to the biosphere. (author)

  2. Development and characterization of new high-level waste form containing LiCl KCl eutectic salts for achieving waste minimization from pyroprocessing

    International Nuclear Information System (INIS)

    Cho, Yong Zun; Kim, In Tae; Park, Hwan Seo; Ahn, Byeung Gil; Eun, Hee Chul; Son, Seock Mo; Ah, Su Na

    2011-12-01

    The purpose of this project is to develop new high level waste (HLW) forms and fabrication processes to dispose of active metal fission products that are removed from electrorefiner salts in the pyroprocessing based fuel cycle. The current technology for disposing of active metal fission products in pyroprocessing involves non selectively discarding of fission product loaded salt in a glass-bonded sodalite ceramic waste form. Selective removal of fission products from the molten salt would greatly minimize the amount of HLW generated and methods were developed to achieve selective separation of fission products during a previous I NERI research project (I NERI 2006 002 K). This I NERI project proceeds from the previous project with the development of suitable waste forms to immobilize the separated fission products. The Korea Atomic Energy Research Institute (KAERI) has focused primarily on developing these waste forms using surrogate waste materials, while the Idaho National Laboratory (INL) has demonstrated fabrication of these waste forms using radioactive electrorefiner salts in hot cell facilities available at INL. Testing and characterization of these radioactive materials was also performed to determine the physical, chemical, and durability properties of the waste forms

  3. Effects of post-disposal gas generation in a repository for spent fuel, high-level waste and long-lived intermediate level waste sited in opalinus clay

    International Nuclear Information System (INIS)

    Johnson, L.; Marschall, P.; Zuidema, P.; Gribi, P.

    2004-07-01

    This comprehensive report issued by the Swiss National Cooperative for the Disposal of Radioactive Waste NAGRA takes a look at post-disposal gas generation in a repository for spent fuel and highly radioactive wastes in Opalinus clay strata. This study provides a comprehensive treatment of the issue of gas generation in a repository for spent fuel (SF), vitrified high-level waste (HLW) and long-lived intermediate-level waste (ILW), sited in the Opalinus clay of the Zuercher Weinland in northern Switzerland. The issue of how gas generation in and transport from waste repositories may influence disposal system performance has been under study for many years, both at Nagra and internationally. The report consists of three main parts: (i) A synthesis of basic information on the host rock and on details of repository construction; (ii) A discussion on gas transport characteristics of the engineered barrier system and the geosphere; (iii) A discussion on the effects of gas on system performance, based on the available information on gas generation, gas transport properties and gas pathways provided in the previous parts of the report. Simplified model calculations based on a mass balance approach for the gas generated within the repository are presented and discussed

  4. The Effects of Oxygen Partial Pressure on Liquidus Temperature of a High-Level Waste Glass with Spinel as the Primary Phase

    International Nuclear Information System (INIS)

    Izak, Pavel; Hrma, Pavel R.; Wilson, Benjamin K.; Vienna, John D.

    2000-01-01

    The redox state of iron affects spinal crystallization in vitrified high-level waste (HLW) glass. Simulated HLW glass with spinel as the primary crystalline phase field was heat treated at constant temperatures within the interval from 850 C to 1300 C under varying atmospheres with oxygen partial pressure, Po2, ranging from 1x10-16 kPa (pure CO) to 101 kPa (pure O2). Liquidus temperature (TL) of glass increased with decreasing Po2 up to Po2 > 3 x 10-9 kPa. At Po2 < 3 x 10-9 kPa, Ni-Fe alloy precipitated from the glass, and TL decreased. Samples were analyzed with optical microscope and scanning electron microscope. The mass fraction of spinel in glass was determined using quantitative X-ray diffraction. Spinel composition was investigated with energy disperse spectroscopy. Ferrous-ferric equilibrium at TL was calculated in a HLW glass as a function of temperature and Po2, based on the previous studies by Schreiber. TL/FeO over the interval 0.0063 < gFeO < 0.051 (1x10-2 kPa < Po2 < 3x10-9 kPa) was estimated from calculated ferrous-ferric equilibrium at TL as 1835 C

  5. End of FY10 report - used fuel disposition technical bases and lessons learned : legal and regulatory framework for high-level waste disposition in the United States.

    Energy Technology Data Exchange (ETDEWEB)

    Weiner, Ruth F.; Blink, James A. (Lawrence Livermore National Laboratory, Livermore, CA); Rechard, Robert Paul; Perry, Frank (Los Alamos National Laboratory, Los Alamos, NM); Jenkins-Smith, Hank C. (University of Oklahoma, Norman, OK); Carter, Joe (Savannah River Nuclear Solutions, Aiken, SC); Nutt, Mark (Argonne National Laboratory, Argonne, IL); Cotton, Tom (Complex Systems Group, Washington DC)

    2010-09-01

    This report examines the current policy, legal, and regulatory framework pertaining to used nuclear fuel and high level waste management in the United States. The goal is to identify potential changes that if made could add flexibility and possibly improve the chances of successfully implementing technical aspects of a nuclear waste policy. Experience suggests that the regulatory framework should be established prior to initiating future repository development. Concerning specifics of the regulatory framework, reasonable expectation as the standard of proof was successfully implemented and could be retained in the future; yet, the current classification system for radioactive waste, including hazardous constituents, warrants reexamination. Whether or not consideration of multiple sites are considered simultaneously in the future, inclusion of mechanisms such as deliberate use of performance assessment to manage site characterization would be wise. Because of experience gained here and abroad, diversity of geologic media is not particularly necessary as a criterion in site selection guidelines for multiple sites. Stepwise development of the repository program that includes flexibility also warrants serious consideration. Furthermore, integration of the waste management system from storage, transportation, and disposition, should be examined and would be facilitated by integration of the legal and regulatory framework. Finally, in order to enhance acceptability of future repository development, the national policy should be cognizant of those policy and technical attributes that enhance initial acceptance, and those policy and technical attributes that maintain and broaden credibility.

  6. Comments on a letter by George D. DeBuchananne (US Geological Survey) regarding the use of salt domes for high-level waste disposal

    International Nuclear Information System (INIS)

    1984-08-01

    The US Geological Survey (USGS) concluded in a letter to the US Department of Energy, dated March 7, 1981, that subsurface geologic conditions in bedded salt are more predictable and less complex than those in domal salt. This predictability is equated with the relative suitability of bedded and domal salt as repository host media. This report comments on the USGS letter. The key points made are as follows: Complexities which may exist in the geologic setting of a salt dome (or other potential host medium) should not a priori preclude the dome from being an acceptable host medium for a high-level waste (HLW) repository. Predictability, as used by the USGS, focused on the spatial extrapolation of information on geologic conditions and should not be confused with predicting the performance of a repository. Notwithstanding the general characteristics of bedded and domal salt, there are salt domes whose individual characteristics should make them as acceptable as potential bedded salt areas for HLW repository sites. Complexities which may occur in the geologic setting of a salt dome can be explored and characterized with sufficient accuracy by available techniques

  7. ICPP tank farm closure study. Volume 2: Engineering design files

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-02-01

    Volume 2 contains the following topical sections: Tank farm heel flushing/pH adjustment; Grouting experiments for immobilization of tank farm heel; Savannah River high level waste tank 20 closure; Tank farm closure information; Clean closure of tank farm; Remediation issues; Remote demolition techniques; Decision concerning EIS for debris treatment facility; CERCLA/RCRA issues; Area of contamination determination; Containment building of debris treatment facility; Double containment issues; Characterization costs; Packaging and disposal options for the waste resulting from the total removal of the tank farm; Take-off calculations for the total removal of soils and structures at the tank farm; Vessel off-gas systems; Jet-grouted polymer and subsurface walls; Exposure calculations for total removal of tank farm; Recommended instrumentation during retrieval operations; High level waste tank concrete encasement evaluation; Recommended heavy equipment and sizing equipment for total removal activities; Tank buoyancy constraints; Grout and concrete formulas for tank heel solidification; Tank heel pH requirements; Tank cooling water; Evaluation of conservatism of vehicle loading on vaults; Typical vault dimensions and approximately tank and vault void volumes; Radiological concerns for temporary vessel off-gas system; Flushing calculations for tank heels; Grout lift depth analysis; Decontamination solution for waste transfer piping; Grout lift determination for filling tank and vault voids; sprung structure vendor data; Grout flow properties through a 2--4 inch pipe; Tank farm load limitations; NRC low level waste grout; Project data sheet calculations; Dose rates for tank farm closure tasks; Exposure and shielding calculations for grout lines; TFF radionuclide release rates; Documentation of the clean closure of a system with listed waste discharge; and Documentation of the ORNL method of radionuclide concentrations in tanks.

  8. ICPP tank farm closure study. Volume 2: Engineering design files

    International Nuclear Information System (INIS)

    1998-02-01

    Volume 2 contains the following topical sections: Tank farm heel flushing/pH adjustment; Grouting experiments for immobilization of tank farm heel; Savannah River high level waste tank 20 closure; Tank farm closure information; Clean closure of tank farm; Remediation issues; Remote demolition techniques; Decision concerning EIS for debris treatment facility; CERCLA/RCRA issues; Area of contamination determination; Containment building of debris treatment facility; Double containment issues; Characterization costs; Packaging and disposal options for the waste resulting from the total removal of the tank farm; Take-off calculations for the total removal of soils and structures at the tank farm; Vessel off-gas systems; Jet-grouted polymer and subsurface walls; Exposure calculations for total removal of tank farm; Recommended instrumentation during retrieval operations; High level waste tank concrete encasement evaluation; Recommended heavy equipment and sizing equipment for total removal activities; Tank buoyancy constraints; Grout and concrete formulas for tank heel solidification; Tank heel pH requirements; Tank cooling water; Evaluation of conservatism of vehicle loading on vaults; Typical vault dimensions and approximately tank and vault void volumes; Radiological concerns for temporary vessel off-gas system; Flushing calculations for tank heels; Grout lift depth analysis; Decontamination solution for waste transfer piping; Grout lift determination for filling tank and vault voids; sprung structure vendor data; Grout flow properties through a 2--4 inch pipe; Tank farm load limitations; NRC low level waste grout; Project data sheet calculations; Dose rates for tank farm closure tasks; Exposure and shielding calculations for grout lines; TFF radionuclide release rates; Documentation of the clean closure of a system with listed waste discharge; and Documentation of the ORNL method of radionuclide concentrations in tanks

  9. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM 2010

    Energy Technology Data Exchange (ETDEWEB)

    West, B.; Waltz, R.

    2011-06-23

    Aqueous radioactive wastes from Savannah River Site (SRS) separations and vitrification processes are contained in large underground carbon steel tanks. Inspections made during 2010 to evaluate these vessels and other waste handling facilities along with evaluations based on data from previous inspections are the subject of this report. The 2010 inspection program revealed that the structural integrity and waste confinement capability of the Savannah River Site waste tanks were maintained. All inspections scheduled per SRR-LWE-2009-00138, HLW Tank Farm Inspection Plan for 2010, were completed. Ultrasonic measurements (UT) performed in 2010 met the requirements of C-ESG-00006, In-Service Inspection Program for High Level Waste Tanks, Rev. 3, and WSRC-TR-2002-00061, Rev.6. UT inspections were performed on Tanks 30, 31 and 32 and the findings are documented in SRNL-STI-2010-00533, Tank Inspection NDE Results for Fiscal Year 2010, Waste Tanks 30, 31 and 32. A total of 5824 photographs were made and 1087 visual and video inspections were performed during 2010. Ten new leaksites at Tank 5 were identified in 2010. The locations of these leaksites are documented in C-ESR-G-00003, SRS High Level Waste Tank Leaksite Information, Rev.5. Ten leaksites at Tank 5 were documented during tank wall/annulus cleaning activities. None of these new leaksites resulted in a release to the environment. The leaksites were documented during wall cleaning activities and the waste nodules associated with the leaksites were washed away. Previously documented leaksites were reactivated at Tank 12 during waste removal activities.

  10. Program plan for evaluation and remediation of the generation and release of flammable gases in Hanford Site waste tanks

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, G.D. (comp.)

    1991-08-01

    This program plan describes the activities being conducted for the resolution of the flammable gas problem that is associated with 23 high-level waste tanks at the Hanford Site. The classification of the wastes in all of these tanks is not final and some wastes may not be high-level wastes. However, until the characterization and classification is complete, all the tanks are treated as if they contain high-level waste. Of the 23 tanks, Tank 241-SY-101 (referred to as Tank 101-SY) has exhibited significant episodic releases of flammable gases (hydrogen and nitrous oxide) for the past 10 years. The major near-term focus of this program is for the understanding and stabilization of this tank. An understanding of the mechanism for gas generation and the processes for the episodic release will be obtained through sampling of the tank contents, laboratory studies, and modeling of the tank behavior. Additional information will be obtained through new and upgraded instrumentation for the tank. A number of remediation, or stabilization, concepts will be evaluated for near-term (2 to 3 years) applications to Tank 101-SY. Detailed safety assessments are required for all activities that will occur in the tank (sampling, removal of equipment, and addition of new instruments). This program plan presents a discussion of each task, provides schedules for near-term activities, and gives a summary of the expected work for fiscal years 1991, 1992, and 1993. 16 refs., 7 figs., 8 tabs.

  11. Program plan for evaluation and remediation of the generation and release of flammable gases in Hanford Site waste tanks

    International Nuclear Information System (INIS)

    Johnson, G.D.

    1991-08-01

    This program plan describes the activities being conducted for the resolution of the flammable gas problem that is associated with 23 high-level waste tanks at the Hanford Site. The classification of the wastes in all of these tanks is not final and some wastes may not be high-level wastes. However, until the characterization and classification is complete, all the tanks are treated as if they contain high-level waste. Of the 23 tanks, Tank 241-SY-101 (referred to as Tank 101-SY) has exhibited significant episodic releases of flammable gases (hydrogen and nitrous oxide) for the past 10 years. The major near-term focus of this program is for the understanding and stabilization of this tank. An understanding of the mechanism for gas generation and the processes for the episodic release will be obtained through sampling of the tank contents, laboratory studies, and modeling of the tank behavior. Additional information will be obtained through new and upgraded instrumentation for the tank. A number of remediation, or stabilization, concepts will be evaluated for near-term (2 to 3 years) applications to Tank 101-SY. Detailed safety assessments are required for all activities that will occur in the tank (sampling, removal of equipment, and addition of new instruments). This program plan presents a discussion of each task, provides schedules for near-term activities, and gives a summary of the expected work for fiscal years 1991, 1992, and 1993. 16 refs., 7 figs., 8 tabs

  12. Hanford ferrocyanide waste chemistry and reactivity preliminary catalyst and initiator screening studies

    International Nuclear Information System (INIS)

    Scheele, R.D.; Bryan, S.A.; Johnston, J.W.; Tingey, J.M.; Burger, L.L.; Hallen, R.T.

    1992-05-01

    During the 1950s, ferrocyanide was used to scavenge radiocesium from aqueous nitrate-containing Hanford wastes. During the production of defense materials and while these wastes were stored in high-level waste tanks at the Hanford Site, some of these wastes were likely mixed with other waste constituents and materials. Recently, Pacific Northwest Laboratory (PNL) was commissioned by Westinghouse Hanford Company (WHC) to investigate the chemical reactivity of these ferrocyanide-bearing wastes. Because of known or potential thermal reactivity hazards associated with ferrocyanide- and nitrate-bearing wastes, and because of the potential for different materials to act as catalysts or initiators of the reactions about which there is concern, we at PNL have begun investigating the effects of the other potential waste constituents. This report presents the results of a preliminary screening study to identify classes of materials that might be in the Hanford high-level waste tanks and that could accelerate or reduce the starting temperature of the reaction(s) of concern. We plan to use the resulted of this study to determine which materials or class of materials merit additional research

  13. 45-Day safety screen results for Tank 241-U-201, push mode, cores 70, 73 and 74

    International Nuclear Information System (INIS)

    Sathyanarayana, P.

    1995-01-01

    Three core samples, each having two segments, from Tank 241-U-201 (U-201) were received by the 222-S Laboratories. Safety screening analysis, such as differential scanning calorimetry (DSC), thermogravimetric analysis (TGA), and total alpha activity were conducted on Core 70, Segment 1 and 2 and on Core 73, Segment 1 and 2. Core 74, Segment 1 and 2 were taken to test rotary bit in push mode sampling. No analysis was requested on Core 74, Segment 1 and 2. Analytical results for the TGA analyses for Core 70, Segment 1, Upper half solid sample was less than the safety screening notification limit of 17 percent water. Notification was made on April 27, 1995. No exotherm was associated with this sample. Analytical results are presented in Tables 1 to 4, with the applicable notification limits shaded

  14. Update to Assessment of Direct Disposal in Unsaturated Tuff of Spent Nuclear Fuel and High-Level Waste Owned by U.S. Department of Energy

    Energy Technology Data Exchange (ETDEWEB)

    P. D. Wheatley (INEEL POC); R. P. Rechard (SNL)

    1998-09-01

    The overall purpose of this study is to provide information and guidance to the Office of Environmental Management of the U.S. Department of Energy (DOE) about the level of characterization necessary to dispose of DOE-owned spent nuclear fuel (SNF). The disposal option modeled was codisposal of DOE SNF with defense high-level waste (DHLW). A specific goal was to demonstrate the influence of DOE SNF, expected to be minor, in a predominately commercial repository using modeling conditions similar to those currently assumed by the Yucca Mountain Project (YMP). A performance assessment (PA) was chosen as the method of analysis. The performance metric for this analysis (referred to as the 1997 PA) was dose to an individual; the time period of interest was 100,000 yr. Results indicated that cumulative releases of 99Tc and 237Np (primary contributors to human dose) from commercial SNF exceed those of DOE SNF both on a per MTHM and per package basis. Thus, if commercial SNF can meet regulatory performance criteria for dose to an individual, then the DOE SNF can also meet the criteria. This result is due in large part to lower burnup of the DOE SNF (less time for irradiation) and to the DOE SNF's small percentage of the total activity (1.5%) and mass (3.8%) of waste in the potential repository. Consistent with the analyses performed for the YMP, the 1997 PA assumed all cladding as failed, which also contributed to the relatively poor performance of commercial SNF compared to DOE SNF.

  15. MEASUREMENT AND CALCULATION OF RADIONUCLIDE ACTIVITIES IN SAVANNAH RIVER SITE HIGH LEVEL WASTE SLUDGE FOR ACCEPTANCE OF DEFENSE WASTE PROCESSING FACILITY GLASS IN A FEDERAL REPOSITORY

    Energy Technology Data Exchange (ETDEWEB)

    Bannochie, C; David Diprete, D; Ned Bibler, N

    2008-12-31

    This paper describes the results of the analyses of High Level Waste (HLW) sludge slurry samples and of the calculations necessary to decay the radionuclides to meet the reporting requirement in the Waste Acceptance Product Specifications (WAPS) [1]. The concentrations of 45 radionuclides were measured. The results of these analyses provide input for radioactive decay calculations used to project the radionuclide inventory at the specified index years, 2015 and 3115. This information is necessary to complete the Production Records at Savannah River Site's Defense Waste Processing Facility (DWPF) so that the final glass product resulting from Macrobatch 5 (MB5) can eventually be submitted to a Federal Repository. Five of the necessary input radionuclides for the decay calculations could not be measured directly due to their low concentrations and/or analytical interferences. These isotopes are Nb-93m, Pd-107, Cd-113m, Cs-135, and Cm-248. Methods for calculating these species from concentrations of appropriate other radionuclides will be discussed. Also the average age of the MB5 HLW had to be calculated from decay of Sr-90 in order to predict the initial concentration of Nb-93m. As a result of the measurements and calculations, thirty-one WAPS reportable radioactive isotopes were identified for MB5. The total activity of MB5 sludge solids will decrease from 1.6E+04 {micro}Ci (1 {micro}Ci = 3.7E+04 Bq) per gram of total solids in 2008 to 2.3E+01 {micro}Ci per gram of total solids in 3115, a decrease of approximately 700 fold. Finally, evidence will be given for the low observed concentrations of the radionuclides Tc-99, I-129, and Sm-151 in the HLW sludges. These radionuclides were reduced in the MB5 sludge slurry to a fraction of their expected production levels due to SRS processing conditions.

  16. Update to Assessment of Direct Disposal in Unsaturated Tuff of Spent Nuclear Fuel and High-Level Waste Owned by U.S. Department of Energy

    International Nuclear Information System (INIS)

    Wheatley, P.D.; Rechard, R.P.

    1998-01-01

    The overall purpose of this study is to provide information and guidance to the Office of Environmental Management of the U.S. Department of Energy (DOE) about the level of characterization necessary to dispose of DOE-owned spent nuclear fuel (SNF). The disposal option modeled was codisposal of DOE SNF with defense high-level waste (DHLW). A specific goal was to demonstrate the influence of DOE SNF, expected to be minor, in a predominately commercial repository using modeling conditions similar to those currently assumed by the Yucca Mountain Project (YMP). A performance assessment (PA) was chosen as the method of analysis. The performance metric for this analysis (referred to as the 1997 PA) was dose to an individual; the time period of interest was 100,000 yr. Results indicated that cumulative releases of 99Tc and 237Np (primary contributors to human dose) from commercial SNF exceed those of DOE SNF both on a per MTHM and per package basis. Thus, if commercial SNF can meet regulatory performance criteria for dose to an individual, then the DOE SNF can also meet the criteria. This result is due in large part to lower burnup of the DOE SNF (less time for irradiation) and to the DOE SNF's small percentage of the total activity (1.5%) and mass (3.8%) of waste in the potential repository. Consistent with the analyses performed for the YMP, the 1997 PA assumed all cladding as failed, which also contributed to the relatively poor performance of commercial SNF compared to DOE SNF

  17. I-NERI-2007-004-K, DEVELOPMENT AND CHARACTERIZATION OF NEW HIGH-LEVEL WASTE FORMS FOR ACHIEVING WASTE MINIMIZATION FROM PYROPROCESSING

    Energy Technology Data Exchange (ETDEWEB)

    S.M. Frank

    2011-09-01

    Work describe in this report represents the final year activities for the 3-year International Nuclear Energy Research Initiative (I-NERI) project: Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing. Used electrorefiner salt that contained actinide chlorides and was highly loaded with surrogate fission products was processed into three candidate waste forms. The first waste form, a high-loaded ceramic waste form is a variant to the CWF produced during the treatment of Experimental Breeder Reactor-II used fuel at the Idaho National Laboratory (INL). The two other waste forms were developed by researchers at the Korean Atomic Energy Research Institute (KAERI). These materials are based on a silica-alumina-phosphate matrix and a zinc/titanium oxide matrix. The proposed waste forms, and the processes to fabricate them, were designed to immobilize spent electrorefiner chloride salts containing alkali, alkaline earth, lanthanide, and halide fission products that accumulate in the salt during the processing of used nuclear fuel. This aspect of the I-NERI project was to demonstrate 'hot cell' fabrication and characterization of the proposed waste forms. The outline of the report includes the processing of the spent electrorefiner salt and the fabrication of each of the three waste forms. Also described is the characterization of the waste forms, and chemical durability testing of the material. While waste form fabrication and sample preparation for characterization must be accomplished in a radiological hot cell facility due to hazardous radioactivity levels, smaller quantities of each waste form were removed from the hot cell to perform various analyses. Characterization included density measurement, elemental analysis, x-ray diffraction, scanning electron microscopy and the Product Consistency Test, which is a leaching method to measure chemical durability. Favorable results from this