Science.gov

Sample records for high level wastes

  1. High level nuclear waste

    SciTech Connect

    Crandall, J L

    1980-01-01

    The DOE Division of Waste Products through a lead office at Savannah River is developing a program to immobilize all US high-level nuclear waste for terminal disposal. DOE high-level wastes include those at the Hanford Plant, the Idaho Chemical Processing Plant, and the Savannah River Plant. Commercial high-level wastes, for which DOE is also developing immobilization technology, include those at the Nuclear Fuel Services Plant and any future commercial fuels reprocessing plants. The first immobilization plant is to be the Defense Waste Processing Facility at Savannah River, scheduled for 1983 project submission to Congress and 1989 operation. Waste forms are still being selected for this plant. Borosilicate glass is currently the reference form, but alternate candidates include concretes, calcines, other glasses, ceramics, and matrix forms.

  2. High-Level Radioactive Waste.

    ERIC Educational Resources Information Center

    Hayden, Howard C.

    1995-01-01

    Presents a method to calculate the amount of high-level radioactive waste by taking into consideration the following factors: the fission process that yields the waste, identification of the waste, the energy required to run a 1-GWe plant for one year, and the uranium mass required to produce that energy. Briefly discusses waste disposal and…

  3. High-level-waste immobilization

    SciTech Connect

    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.

  4. Optimizing High Level Waste Disposal

    SciTech Connect

    Dirk Gombert

    2005-09-01

    If society is ever to reap the potential benefits of nuclear energy, technologists must close the fuel-cycle completely. A closed cycle equates to a continued supply of fuel and safe reactors, but also reliable and comprehensive closure of waste issues. High level waste (HLW) disposal in borosilicate glass (BSG) is based on 1970s era evaluations. This host matrix is very adaptable to sequestering a wide variety of radionuclides found in raffinates from spent fuel reprocessing. However, it is now known that the current system is far from optimal for disposal of the diverse HLW streams, and proven alternatives are available to reduce costs by billions of dollars. The basis for HLW disposal should be reassessed to consider extensive waste form and process technology research and development efforts, which have been conducted by the United States Department of Energy (USDOE), international agencies and the private sector. Matching the waste form to the waste chemistry and using currently available technology could increase the waste content in waste forms to 50% or more and double processing rates. Optimization of the HLW disposal system would accelerate HLW disposition and increase repository capacity. This does not necessarily require developing new waste forms, the emphasis should be on qualifying existing matrices to demonstrate protection equal to or better than the baseline glass performance. Also, this proposed effort does not necessarily require developing new technology concepts. The emphasis is on demonstrating existing technology that is clearly better (reliability, productivity, cost) than current technology, and justifying its use in future facilities or retrofitted facilities. Higher waste processing and disposal efficiency can be realized by performing the engineering analyses and trade-studies necessary to select the most efficient methods for processing the full spectrum of wastes across the nuclear complex. This paper will describe technologies being

  5. High-level waste processing and disposal

    NASA Astrophysics Data System (ADS)

    Crandall, J. L.; Drause, H.; Sombret, C.; Uematsu, K.

    The national high level waste disposal plans for France, the Federal Republic of Germany, Japan, and the United States are covered. Three conclusions are reached. The first conclusion is that an excellent technology already exists for high level waste disposal. With appropriate packaging, spent fuel seems to be an acceptable waste form. Borosilicate glass reprocessing waste forms are well understood, in production in France, and scheduled for production in the next few years in a number of other countries. For final disposal, a number of candidate geological repository sites have been identified and several demonstration sites opened. The second conclusion is that adequate financing and a legal basis for waste disposal are in place in most countries. Costs of high level waste disposal will probably and about 5 to 10% to the costs of nuclear electric power. Third conclusion is less optimistic.

  6. PAIRWISE BLENDING OF HIGH LEVEL WASTE (HLW)

    SciTech Connect

    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.

  7. High-level waste qualification: Managing uncertainty

    SciTech Connect

    Pulsipher, B.A.

    1993-09-01

    A vitrification facility is being developed by the U.S. Department of Energy (DOE) at the West Valley Demonstration Plant (WVDP) near Buffalo, New York, where approximately 300 canisters of high-level nuclear waste glass will be produced. To assure that the produced waste form is acceptable, uncertainty must be managed. Statistical issues arise due to sampling, waste variations, processing uncertainties, and analytical variations. This paper presents elements of a strategy to characterize and manage the uncertainties associated with demonstrating that an acceptable waste form product is achieved. Specific examples are provided within the context of statistical work performed by Pacific Northwest Laboratory (PNL).

  8. High-Level Waste Melter Study Report

    SciTech Connect

    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. High-level radioactive wastes. Supplement 1

    SciTech Connect

    McLaren, L.H.

    1984-09-01

    This bibliography contains information on high-level radioactive wastes included in the Department of Energy's Energy Data Base from August 1982 through December 1983. These citations are to research reports, journal articles, books, patents, theses, and conference papers from worldwide sources. Five indexes, each preceded by a brief description, are provided: Corporate Author, Personal Author, Subject, Contract Number, and Report Number. 1452 citations.

  10. Technetium Chemistry in High-Level Waste

    SciTech Connect

    Hess, Nancy J.

    2006-06-01

    Tc contamination is found within the DOE complex at those sites whose mission involved extraction of plutonium from irradiated uranium fuel or isotopic enrichment of uranium. At the Hanford Site, chemical separations and extraction processes generated large amounts of high level and transuranic wastes that are currently stored in underground tanks. The waste from these extraction processes is currently stored in underground High Level Waste (HLW) tanks. However, the chemistry of the HLW in any given tank is greatly complicated by repeated efforts to reduce volume and recover isotopes. These processes ultimately resulted in mixing of waste streams from different processes. As a result, the chemistry and the fate of Tc in HLW tanks are not well understood. This lack of understanding has been made evident in the failed efforts to leach Tc from sludge and to remove Tc from supernatants prior to immobilization. Although recent interest in Tc chemistry has shifted from pretreatment chemistry to waste residuals, both needs are served by a fundamental understanding of Tc chemistry.

  11. 40 CFR 227.30 - High-level radioactive waste.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 40 Protection of Environment 25 2011-07-01 2011-07-01 false High-level radioactive waste. 227.30 Section 227.30 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) OCEAN DUMPING...-level radioactive waste. High-level radioactive waste means the aqueous waste resulting from...

  12. 40 CFR 227.30 - High-level radioactive waste.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 40 Protection of Environment 25 2014-07-01 2014-07-01 false High-level radioactive waste. 227.30 Section 227.30 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) OCEAN DUMPING...-level radioactive waste. High-level radioactive waste means the aqueous waste resulting from...

  13. 40 CFR 227.30 - High-level radioactive waste.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 40 Protection of Environment 26 2012-07-01 2011-07-01 true High-level radioactive waste. 227.30 Section 227.30 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) OCEAN DUMPING...-level radioactive waste. High-level radioactive waste means the aqueous waste resulting from...

  14. 40 CFR 227.30 - High-level radioactive waste.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 40 Protection of Environment 26 2013-07-01 2013-07-01 false High-level radioactive waste. 227.30 Section 227.30 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) OCEAN DUMPING...-level radioactive waste. High-level radioactive waste means the aqueous waste resulting from...

  15. DEFENSE HIGH LEVEL WASTE GLASS DEGRADATION

    SciTech Connect

    W. Ebert

    2001-09-20

    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&O 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. Multipurpose optimization models for high level waste vitrification

    SciTech Connect

    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.

  17. Decontamination of high-level waste canisters

    SciTech Connect

    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.

  18. Decision Document for Heat Removal from High Level Waste Tanks

    SciTech Connect

    WILLIS, W.L.

    2000-07-31

    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.

  19. Reference commercial high-level waste glass and canister definition.

    SciTech Connect

    Slate, S.C.; Ross, W.A.; Partain, W.L.

    1981-09-01

    This report presents technical data and performance characteristics of a high-level waste glass and canister intended for use in the design of a complete waste encapsulation package suitable for disposal in a geologic repository. The borosilicate glass contained in the stainless steel canister represents the probable type of high-level waste product that will be produced in a commercial nuclear-fuel reprocessing plant. Development history is summarized for high-level liquid waste compositions, waste glass composition and characteristics, and canister design. The decay histories of the fission products and actinides (plus daughters) calculated by the ORIGEN-II code are presented.

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

    SciTech Connect

    Stegen, G.E.

    1996-02-13

    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.

  1. Handbook of high-level radioactive waste transportation

    SciTech Connect

    Sattler, L.R.

    1992-10-01

    The High-Level Radioactive Waste Transportation Handbook serves as a reference to which state officials and members of the general public may turn for information on radioactive waste transportation and on the federal government`s system for transporting this waste under the Civilian Radioactive Waste Management Program. The Handbook condenses and updates information contained in the Midwestern High-Level Radioactive Waste Transportation Primer. It is intended primarily to assist legislators who, in the future, may be called upon to enact legislation pertaining to the transportation of radioactive waste through their jurisdictions. The Handbook is divided into two sections. The first section places the federal government`s program for transporting radioactive waste in context. It provides background information on nuclear waste production in the United States and traces the emergence of federal policy for disposing of radioactive waste. The second section covers the history of radioactive waste transportation; summarizes major pieces of legislation pertaining to the transportation of radioactive waste; and provides an overview of the radioactive waste transportation program developed by the US Department of Energy (DOE). To supplement this information, a summary of pertinent federal and state legislation and a glossary of terms are included as appendices, as is a list of publications produced by the Midwestern Office of The Council of State Governments (CSG-MW) as part of the Midwestern High-Level Radioactive Waste Transportation Project.

  2. Process for solidifying high-level nuclear waste

    DOEpatents

    Ross, Wayne A.

    1978-01-01

    The addition of a small amount of reducing agent to a mixture of a high-level radioactive waste calcine and glass frit before the mixture is melted will produce a more homogeneous glass which is leach-resistant and suitable for long-term storage of high-level radioactive waste products.

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

    SciTech Connect

    Bernadzikowski, T. A.; Allender, J. S.; Butler, J. L.; Gordon, D. E.; Gould, Jr., T. H.; Stone, J. A.

    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.

  4. FLUIDIZED BED STEAM REFORMING ENABLING ORGANIC HIGH LEVEL WASTE DISPOSAL

    SciTech Connect

    Williams, M

    2008-05-09

    Waste streams planned for generation by the Global Nuclear Energy Partnership (GNEP) and existing radioactive High Level Waste (HLW) streams containing organic compounds such as the Tank 48H waste stream at Savannah River Site have completed simulant and radioactive testing, respectfully, by Savannah River National Laboratory (SRNL). GNEP waste streams will include up to 53 wt% organic compounds and nitrates up to 56 wt%. Decomposition of high nitrate streams requires reducing conditions, e.g. provided by organic additives such as sugar or coal, to reduce NOX in the off-gas to N2 to meet Clean Air Act (CAA) standards during processing. Thus, organics will be present during the waste form stabilization process regardless of the GNEP processes utilized and exists in some of the high level radioactive waste tanks at Savannah River Site and Hanford Tank Farms, e.g. organics in the feed or organics used for nitrate destruction. Waste streams containing high organic concentrations cannot be stabilized with the existing HLW Best Developed Available Technology (BDAT) which is HLW vitrification (HLVIT) unless the organics are removed by pretreatment. The alternative waste stabilization pretreatment process of Fluidized Bed Steam Reforming (FBSR) operates at moderate temperatures (650-750 C) compared to vitrification (1150-1300 C). The FBSR process has been demonstrated on GNEP simulated waste and radioactive waste containing high organics from Tank 48H to convert organics to CAA compliant gases, create no secondary liquid waste streams and create a stable mineral waste form.

  5. Stability of High-Level Waste Forms

    SciTech Connect

    Besmann, Theodore M.; Vienna, John D.

    2005-09-30

    The objective of the proposed effort is to use a new approach to develop solution models of complex waste glass systems and spent fuel that are predictive with regard to composition, phase separation, and volatility. The effort will also yield thermodynamic values for waste components that are fundamentally required for corrosion models used to predict the leaching/corrosion behavior for waste glass and spent fuel material. This basic information and understanding of chemical behavior can subsequently be used directly in computational models of leaching and transport in geologic media, in designing and engineering waste forms and barrier systems, and in prediction of chemical interactions.

  6. STABILITY OF HIGH-LEVEL WASTE FORMS

    EPA Science Inventory

    The assessment of release of radionuclides from waste repositories substantially depends on the leaching behavior of the spent fuel or waste form. Assumed rates based on dissolution of specific phases (assumption of unit activity) will lead to potentially grossly overestimated va...

  7. Final report on cermet high-level waste forms

    SciTech Connect

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

    1981-08-01

    Cermets are being developed as an alternate method for the fixation of defense and commercial high level radioactive waste in a terminal disposal form. Following initial feasibility assessments of this waste form, consisting of ceramic particles dispersed in an iron-nickel base alloy, significantly improved processing methods were developed. The characterization of cermets has continued through property determinations on samples prepared by various methods from a variety of simulated and actual high-level wastes. This report describes the status of development of the cermet waste form as it has evolved since 1977. 6 tables, 18 figures.

  8. High-level waste program progress report, April 1, 1980-June 30, 1980

    SciTech Connect

    1980-08-01

    The highlights of this report are on: waste management analysis for nuclear fuel cycles; fixation of waste in concrete; study of ceramic and cermet waste forms; alternative high-level waste forms development; and high-level waste container development.

  9. High level waste characterization in support of low level waste certification. I. HLW supernate radionuclide characterization

    SciTech Connect

    Jamison, M.E.; d`Entremont, P.D.; Clemmons, J.S.; Bess, C.E.; Brown, D.F.

    1994-07-08

    High Level Waste Programs has radioactive waste storage, treatment and processing facilities that are located in the F and H Areas at the Savannah River Site. These facilities include the Effluent Treatment Facility (ETF), F and H Area Tank Farms, Extended Sludge Processing (ESP), and In-Tank Precipitation (ITP). Job wastes are generated from operation, maintenance, and construction activities inside radiological areas. These items may have been contaminated with radioactive supernate, salt, and sludge material. Most of these wastes will be disposed of in the E-area Vaults. Therefore, an isotopic and hazardous characterization must be performed. The characterization of HLW supernate radionuclides is discussed in Chapter I. The characterization for salt and sludge phases, which can also contaminate LLW, will be included in other Chapters.

  10. Review of high-level waste form properties. [146 bibliographies

    SciTech Connect

    Rusin, J.M.

    1980-12-01

    This report is a review of waste form options for the immobilization of high-level-liquid wastes from the nuclear fuel cycle. This review covers the status of international research and development on waste forms as of May 1979. Although the emphasis in this report is on waste form properties, process parameters are discussed where they may affect final waste form properties. A summary table is provided listing properties of various nuclear waste form options. It is concluded that proposed waste forms have properties falling within a relatively narrow range. In regard to crystalline versus glass waste forms, the conclusion is that either glass of crystalline materials can be shown to have some advantage when a single property is considered; however, at this date no single waste form offers optimum properties over the entire range of characteristics investigated. A long-term effort has been applied to the development of glass and calcine waste forms. Several additional waste forms have enough promise to warrant continued research and development to bring their state of development up to that of glass and calcine. Synthetic minerals, the multibarrier approach with coated particles in a metal matrix, and high pressure-high temperature ceramics offer potential advantages and need further study. Although this report discusses waste form properties, the total waste management system should be considered in the final selection of a waste form option. Canister design, canister materials, overpacks, engineered barriers, and repository characteristics, as well as the waste form, affect the overall performance of a waste management system. These parameters were not considered in this comparison.

  11. International High Level Nuclear Waste Management

    ERIC Educational Resources Information Center

    Dreschhoff, Gisela; And Others

    1974-01-01

    Discusses the radioactive waste management in Belgium, Canada, France, Germany, India, Italy, Japan, the United Kingdom, the United States, and the USSR. Indicates that scientists and statesmen should look beyond their own lifetimes into future centuries and millennia to conduct long-range plans essential to protection of future generations. (CC)

  12. Neptunium estimation in dissolver and high-level-waste solutions

    SciTech Connect

    Pathak, P.N.; Prabhu, D.R.; Kanekar, A.S.; Manchanda, V.K.

    2008-07-01

    This papers deals with the optimization of the experimental conditions for the estimation of {sup 237}Np in spent-fuel dissolver/high-level waste solutions using thenoyltrifluoroacetone as the extractant. (authors)

  13. High-Level Waste System Process Interface Description

    SciTech Connect

    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.

  14. Disposal of high-level nuclear waste in space

    NASA Astrophysics Data System (ADS)

    Coopersmith, Jonathan

    1992-08-01

    A solution of launching high-level nuclear waste into space is suggested. Disposal in space includes solidifying the wastes, embedding them in an explosion-proof vehicle, and launching it into earth orbit, and then into a solar orbit. The benefits of such a system include not only the safe disposal of high-level waste but also the establishment of an infrastructure for large-scale space exploration and development. Particular attention is given to the wide range of technical choices along with the societal, economic, and political factors needed for success.

  15. Solidification of Savannah River Plant high-level waste

    SciTech Connect

    Maher, R; Shafranek, L F; Stevens, III, W R

    1983-01-01

    The Department of Energy, in accord with recommendations from the Du Pont Company, has started construction of a Defense Waste Processing Facility (DWPF) at the Savannah River Plant. The facility should be completed by the end of 1988, and full-scale operation should begin in 1990. This facility will immobilize in borosilicate glass the large quantity of high-level radioactive waste now stored at the plant plus the waste to be generated from continued chemical reprocessing operations. The existing wastes at the Savannah River Plant will be completely converted by about 2010. 21 figures.

  16. RETENTION OF SULFATE IN HIGH LEVEL RADIOACTIVE WASTE GLASS

    SciTech Connect

    Fox, K.

    2010-09-07

    High level radioactive wastes are being vitrified at the Savannah River Site for long term disposal. Many of the wastes contain sulfate at concentrations that can be difficult to retain in borosilicate glass. This study involves efforts to optimize the composition of a glass frit for combination with the waste to improve sulfate retention while meeting other process and product performance constraints. The fabrication and characterization of several series of simulated waste glasses are described. The experiments are detailed chronologically, to provide insight into part of the engineering studies used in developing frit compositions for an operating high level waste vitrification facility. The results lead to the recommendation of a specific frit composition and a concentration limit for sulfate in the glass for the next batch of sludge to be processed at Savannah River.

  17. Defense High Level Waste Disposal Container System Description Document

    SciTech Connect

    N. E. Pettit

    2001-07-13

    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

  18. Overview of high-level waste management accomplishments

    SciTech Connect

    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.

  19. A comparison of high-level waste form characteristics

    SciTech Connect

    Salmon, R.; Notz, K.J.

    1991-01-01

    There are currently about 1055 million curies of high-level waste with a thermal output of about 2950 kilowatts (KW) at four sites in the United States: West Valley Demonstration Project (WVDP), Savannah River Site (SRS), Hanford Site (HANF), and Idaho National Engineering Laboratory (INEL). These quantities are expected to increase to about 1200 million curies and 3570 kw by the end of year 2020. Under the Nuclear Waste Policy Act, this high-level waste must ultimately be disposed of in a geologic repository. Accordingly, canisters of high-level waste immobilized in borosilicate glass or glass-ceramic mixtures are to be produced at the four sites and stored there until a repository becomes available. Data on the estimated production schedules and on the physical, chemical, and radiological characteristics of the canisters of immobilized high-level waste have been collected in OCRWM's Waste Characteristics Data Base, including recent updates an revisions. Comparisons of some of these data for the four sites are presented in this report. 14 refs., 3 tabs.

  20. Waste acceptance product specifications for vitrified high-level waste forms. Revision 1

    SciTech Connect

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

    1993-06-01

    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.{sup 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.

  1. High level radioactive waste glass production and product description

    SciTech Connect

    Sproull, J.F.; Marra, S.L.; Jantzen, C.M.

    1993-12-01

    This report examines borosilicate glass as a means of immobilizing high-level radioactive wastes. Borosilicate glass will encapsulate most of the defense and some of the commercial HLW in the US. The resulting waste forms must meet the requirements of the WA-SRD and the WAPS, which include a short term PCT durability test. The waste form producer must report the composition(s) of the borosilicate waste glass(es) produced but can choose the composition(s) to meet site-specific requirements. Although the waste form composition is the primary determinant of durability, the redox state of the glass; the existence, content, and composition of crystals; and the presence of glass-in-glass phase separation can affect durability. The waste glass should be formulated to avoid phase separation regions. The ultimate result of this effort will be a waste form which is much more stable and potentially less mobile than the liquid high level radioactive waste is currently.

  2. Management of data quality of high level waste characterization

    SciTech Connect

    Winters, W.I., Westinghouse Hanford

    1996-06-12

    Over the past 10 years, the Hanford Site has been transitioning from nuclear materials production to Site cleanup operations. High-level waste characterization at the Hanford Site provides data to support present waste processing operations, tank safety programs, and future waste disposal programs. Quality elements in the high-level waste characterization program will be presented by following a sample through the data quality objective, sampling, laboratory analysis and data review process. Transition from production to cleanup has resulted in changes in quality systems and program; the changes, as well as other issues in these quality programs, will be described. Laboratory assessment through quality control and performance evaluation programs will be described, and data assessments in the laboratory and final reporting in the tank characterization reports will be discussed.

  3. Case for retrievable high-level nuclear waste disposal

    USGS Publications Warehouse

    Roseboom, Eugene H., Jr.

    1994-01-01

    Plans for the nation's first high-level nuclear waste repository have called for permanently closing and sealing the repository soon after it is filled. However, the hydrologic environment of the proposed site at Yucca Mountain, Nevada, should allow the repository to be kept open and the waste retrievable indefinitely. This would allow direct monitoring of the repository and maintain the options for future generations to improve upon the disposal methods or use the uranium in the spent fuel as an energy resource.

  4. Materials Science of High-Level Nuclear Waste Immobilization

    SciTech Connect

    Weber, William J.; Navrotsky, Alexandra; Stefanovsky, S. V.; Vance, E. R.; Vernaz, Etienne Y.

    2009-01-09

    With the increasing demand for the development of more nuclear power comes the responsibility to address the technical challenges of immobilizing high-level nuclear wastes in stable solid forms for interim storage or disposition in geologic repositories. The immobilization of high-level nuclear wastes has been an active area of research and development for over 50 years. Borosilicate glasses and complex ceramic composites have been developed to meet many technical challenges and current needs, although regulatory issues, which vary widely from country to country, have yet to be resolved. Cooperative international programs to develop advanced proliferation-resistant nuclear technologies to close the nuclear fuel cycle and increase the efficiency of nuclear energy production might create new separation waste streams that could demand new concepts and materials for nuclear waste immobilization. This article reviews the current state-of-the-art understanding regarding the materials science of glasses and ceramics for the immobilization of high-level nuclear waste and excess nuclear materials and discusses approaches to address new waste streams.

  5. Corrosion and failure processes in high-level waste tanks

    SciTech Connect

    Mahidhara, R.K.; Elleman, T.S.; Murty, K.L.

    1992-11-01

    A large amount of radioactive waste has been stored safely at the Savannah River and Hanford sites over the past 46 years. The aim of this report is to review the experimental corrosion studies at Savannah River and Hanford with the intention of identifying the types and rates of corrosion encountered and indicate how these data contribute to tank failure predictions. The compositions of the High-Level Wastes, mild steels used in the construction of the waste tanks and degradation-modes particularly stress corrosion cracking and pitting are discussed. Current concerns at the Hanford Site are highlighted.

  6. Talc-silicon glass-ceramic waste forms for immobilization of high- level calcined waste

    SciTech Connect

    Vinjamuri, K.

    1993-06-01

    Talc-silicon glass-ceramic waste forms are being evaluated as candidates for immobilization of the high level calcined waste stored onsite at the Idaho Chemical Processing Plant. These glass-ceramic waste forms were prepared by hot isostatically pressing a mixture of simulated nonradioactive high level calcined waste, talc, silicon and aluminum metal additives. The waste forms were characterized for density, chemical durability, and glass and crystalline phase compositions. The results indicate improved density and chemical durability as the silicon content is increased.

  7. Defense High Level Waste Disposal Container System Description

    SciTech Connect

    2000-10-12

    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. U.S. 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 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 which to identify the disposal container and its contents. Different materials

  8. High Level Waste (HLW) Feed Process Control Strategy

    SciTech Connect

    STAEHR, T.W.

    2000-06-14

    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.

  9. High level radioactive waste management facility design criteria

    SciTech Connect

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

    1993-10-01

    This paper discusses the engineering systems for the structural design of the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). At the DWPF, high level radioactive liquids will be mixed with glass particles and heated in a melter. This molten glass will then be poured into stainless steel canisters where it will harden. This process will transform the high level waste into a more stable, manageable substance. This paper discuss the structural design requirements for this unique one of a kind facility. A special emphasis will be concentrated on the design criteria pertaining to earthquake, wind and tornado, and flooding.

  10. High-level waste management technology program plan

    SciTech Connect

    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.

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

    SciTech Connect

    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.

  12. HIGH ALUMINUM HLW (HIGH LEVEL WASTE ) GLASSES FOR HANFORDS WTP (WASTE TREATMENT PROJECT)

    SciTech Connect

    KRUGER AA; BOWAN BW; JOSEPH I; GAN H; KOT WK; MATLACK KS; PEGG IL

    2010-01-04

    This paper presents the results of glass formulation development and melter testing to identify high waste loading glasses to treat high-Al high level waste (HLW) at Hanford. Previous glass formulations developed for this HLW had high waste loadings but their processing rates were lower that desired. The present work was aimed at improving the glass processing rate while maintaining high waste loadings. Glass formulations were designed, prepared at crucible-scale and characterized to determine their properties relevant to processing and product quality. Glass formulations that met these requirements were screened for melt rates using small-scale tests. The small-scale melt rate screening included vertical gradient furnace (VGF) and direct feed consumption (DFC) melter tests. Based on the results of these tests, modified glass formulations were developed and selected for larger scale melter tests to determine their processing rate. Melter tests were conducted on the DuraMelter 100 (DMIOO) with a melt surface area of 0.11 m{sup 2} and the DuraMelter 1200 (DMI200) HLW Pilot Melter with a melt surface area of 1.2 m{sup 2}. The newly developed glass formulations had waste loadings as high as 50 wt%, with corresponding Al{sub 2}O{sub 3} concentration in the glass of 26.63 wt%. The new glass formulations showed glass production rates as high as 1900 kg/(m{sup 2}.day) under nominal melter operating conditions. The demonstrated glass production rates are much higher than the current requirement of 800 kg/(m{sup 2}.day) and anticipated future enhanced Hanford Tank Waste Treatment and Immobilization Plant (WTP) requirement of 1000 kg/(m{sup 2}.day).

  13. Radioactive high level waste insight modelling for geological disposal facilities

    NASA Astrophysics Data System (ADS)

    Carter, Alexander; Kelly, Martin; Bailey, Lucy

    Within this paper we present a simplified analytical model to provide insight into the key performance measures of a generic disposal system for high level waste within a geological disposal facility. The model assumes a low solubility waste matrix within a corrosion resistant disposal container surrounded by a low permeability buffer. Radionuclides migrate from the disposal area through a porous geosphere to the biosphere and give a radiological dose to a receptor. The system of equations describing the migration is transformed into Laplace space and an approximation used to determine peak values for the radionuclide mass transfer rate entering the biosphere. Results from the model are compared with those from more detailed numerical models for key radionuclides in the UK high level waste inventory. Such an insight model can provide a valuable second line of argument to assist in confirming the results of more detailed models and build confidence in the safety case for a geological disposal facility.

  14. 40 CFR 227.30 - High-level radioactive waste.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 24 2010-07-01 2010-07-01 false High-level radioactive waste. 227.30 Section 227.30 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) OCEAN DUMPING CRITERIA FOR THE EVALUATION OF PERMIT APPLICATIONS FOR OCEAN DUMPING OF MATERIALS Definitions § 227.30...

  15. MIXING PROCESSES IN HIGH-LEVEL WASTE TANKS

    EPA Science Inventory

    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...

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

    SciTech Connect

    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.

  17. Development of Ceramic Waste Forms for High-Level Nuclear Waste Over the Last 30 Years

    SciTech Connect

    Vance, Eric

    2007-07-01

    Many types of ceramics have been put forward for immobilisation of high-level waste (HLW) from reprocessing of nuclear power plant fuel or weapons production. After describing some historical aspects of waste form research, the essential features of the chemical design and processing of these different ceramic types will be discussed briefly. Given acceptable laboratory and long-term predicted performance based on appropriately rigorous chemical design, the important processing parameters are mostly waste loading, waste throughput, footprint, offgas control/minimization, and the need for secondary waste treatment. It is concluded that the 'problem of high-level nuclear waste' is largely solved from a technical point of view, within the current regulatory framework, and that the main remaining question is which technical disposition method is optimum for a given waste. (author)

  18. Application of SYNROC to high-level defense wastes

    SciTech Connect

    Tewhey, J.D.; Hoenig, C.L.; Newkirk, H.W.; Rozsa, R.B.; Coles, D.G.; Ryerson, F.J.

    1981-01-01

    The SYNROC method for immobilization of high-level nuclear reactor wastes is currently being applied to US defense wastes in tank storage at Savannah River, South Carolina. The minerals zirconolite, perovskite, and hollandite are used in SYNROC D formulations to immobilize fission products and actinides that comprise up to 10% of defense waste sludges and coexisting solutions. Additional phases in SYNROC D are nepheline, the host phase for sodium; and spinel, the host for excess aluminum and iron. Up to 70 wt % of calcined sludge can be incorporated with 30 wt % of SYNROC additives to produce a waste form consisting of 10% nepheline, 30% spinel, and approximately 20% each of the radioactive waste-bearing phases. Urea coprecipitation and spray drying/calcining methods have been used in the laboratory to produce homogeneous, reactive ceramic powders. Hot pressing and sintering at temperatures from 1000 to 1100/sup 0/C result in waste form products with greater than 97% of theoretical density. Hot isostatic pressing has recently been implemented as a processing alternative. Characterization of waste-form mineralogy has been done by means of XRD, SEM, and electron microprobe. Leaching of SYNROC D samples is currently being carried out. Assessment of radiation damage effects and physical properties of SYNROC D will commence in FY 81.

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

    SciTech Connect

    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.

  20. Spent fuel and high-level radioactive waste transportation report

    SciTech Connect

    Not Available

    1990-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  1. Spent Fuel and High-Level Radioactive Waste Transportation Report

    SciTech Connect

    Not Available

    1992-03-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by SSEB in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste Issues. In addition. this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  2. Spent fuel and high-level radioactive waste transportation report

    SciTech Connect

    Not Available

    1989-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages sew be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

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

    SciTech Connect

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

    1995-12-31

    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.

  4. Towards increased waste loading in high level waste glasses: Developing a better understanding of crystallization behavior

    DOE PAGESBeta

    Marra, James C.; Kim, Dong -Sang

    2014-12-18

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JCHM) 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. Thus, 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.more » 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 (with higher Al2O3). 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.« less

  5. Towards increased waste loading in high level waste glasses: Developing a better understanding of crystallization behavior

    SciTech Connect

    Marra, James C.; Kim, Dong -Sang

    2014-12-18

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JCHM) 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. Thus, 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 (with higher Al2O3). 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.

  6. Progress of the High Level Waste Program at the Defense Waste Processing Facility - 13178

    SciTech Connect

    Bricker, Jonathan M.; Fellinger, Terri L.; Staub, Aaron V.; Ray, Jeff W.; Iaukea, John F.

    2013-07-01

    The Defense Waste Processing Facility at the Savannah River Site treats and immobilizes High Level Waste into a durable borosilicate glass for safe, permanent storage. The High Level Waste program significantly reduces environmental risks associated with the storage of radioactive waste from legacy efforts to separate fissionable nuclear material from irradiated targets and fuels. In an effort to support the disposition of radioactive waste and accelerate tank closure at the Savannah River Site, the Defense Waste Processing Facility recently implemented facility and flowsheet modifications to improve production by 25%. These improvements, while low in cost, translated to record facility production in fiscal years 2011 and 2012. In addition, significant progress has been accomplished on longer term projects aimed at simplifying and expanding the flexibility of the existing flowsheet in order to accommodate future processing needs and goals. (authors)

  7. Advanced waste form and Melter development for treatment of troublesome high-level wastes

    SciTech Connect

    Marra, James; Kim, Dong -Sang; Maio, Vincent

    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

  8. SIMULANT DEVELOPMENT FOR SAVANNAH RIVER SITE HIGH LEVEL WASTE

    SciTech Connect

    Stone, M; Russell Eibling, R; David Koopman, D; Dan Lambert, D; Paul Burket, P

    2007-09-04

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site vitrifies High Level Waste (HLW) for repository internment. The process consists of three major steps: waste pretreatment, vitrification, and canister decontamination/sealing. The HLW consists of insoluble metal hydroxides (primarily iron, aluminum, magnesium, manganese, and uranium) and soluble sodium salts (carbonate, hydroxide, nitrite, nitrate, and sulfate). The HLW is processed in large batches through DWPF; DWPF has recently completed processing Sludge Batch 3 (SB3) and is currently processing Sludge Batch 4 (SB4). The composition of metal species in SB4 is shown in Table 1 as a function of the ratio of a metal to iron. Simulants remove radioactive species and renormalize the remaining species. Supernate composition is shown in Table 2.

  9. Modeling and optimization of defense high level waste removal sequencing

    NASA Astrophysics Data System (ADS)

    Paul, Pran Krishna

    A novel methodology has been developed which makes possible a very fast running computational tool, capable of performing 30 to 50 years of simulation of the entire Savannah River Site (SRS) high level waste complex in less than 2 minutes on a work station. The methodology has been implemented in the Production Planning Model (ProdMod) simulation code which uses Aspen Technology's dynamic simulation software development package SPEEDUP. ProdMod is a pseudo-dynamic simulation code solely based on algebraic equations, using no differential equations. The dynamic nature of the plant process is captured using linear constructs in which the time dependence is implicit. Another innovative approach implemented in ProdMod development is the mapping of event-space on to time-space and vice versa, which accelerates the computation without sacrificing the necessary details in the event-space. ProdMod uses this approach in coupling the time-space continuous simulation with the event-space batch simulation, avoiding the discontinuities inherent in dynamic simulation batch processing. In addition, a general purpose optimization scheme has been devised based on the pseudo-dynamic constructs and the event- and time-space algorithms of ProdMod. The optimization scheme couples a FORTRAN based stand-alone optimization driver with the SPEEDUP based ProdMod simulator to perform dynamic optimization. The scheme is capable of generating single or multiple optimal input conditions for different types of objective functions over single or multiple years of operations depending on the nature of the objective function and operating constraints. The resultant optimal inputs are then interfaced with ProdMod to simulate the dynamic behavior of the waste processing operations. At the conclusion on an optimized advancement step, the simulation parameters are then passed to the optimization driver to generate the next set of optimized parameters. An optimization algorithm using linear programming

  10. High-level waste program integration within the DOE complex

    SciTech Connect

    Valentine, J.H.; Davis, N.R.; Malone, K.; Schaus, P.S.

    1998-03-01

    Eleven major Department of Energy (DOE) site contractors were chartered by the Assistant Secretary to use a systems engineering approach to develop and evaluate technically defensible cost savings opportunities across the complex. Known as the complex-wide Environmental Management Integration (EMI), this process evaluated all the major DOE waste streams including high level waste (HLW). Across the DOE complex, this waste stream has the highest life cycle cost and is scheduled to take until at least 2035 before all HLW is processed for disposal. Technical contract experts from the four DOE sites that manage high level waste participated in the integration analysis: Hanford, Savannah River Site (SRS), Idaho National Engineering and Environmental Laboratory (INEEL), and West Valley Demonstration Project (WVDP). In addition, subject matter experts from the Yucca Mountain Project and the Tanks Focus Area participated in the analysis. Also, departmental representatives from the US Department of Energy Headquarters (DOE-HQ) monitored the analysis and results. Workouts were held throughout the year to develop recommendations to achieve a complex-wide integrated program. From this effort, the HLW Environmental Management (EM) Team identified a set of programmatic and technical opportunities that could result in potential cost savings and avoidance in excess of $18 billion and an accelerated completion of the HLW mission by seven years. The cost savings, schedule improvements, and volume reduction are attributed to a multifaceted HLW treatment disposal strategy which involves waste pretreatment, standardized waste matrices, risk-based retrieval, early development and deployment of a shipping system for glass canisters, and reasonable, low cost tank closure.

  11. Review of High Level Waste Tanks Ultrasonic Inspection Data

    SciTech Connect

    Wiersma, B

    2006-03-09

    A review of the data collected during ultrasonic inspection of the Type I high level waste tanks has been completed. The data was analyzed for relevance to the possibility of vapor space corrosion and liquid/air interface corrosion. The review of the Type I tank UT inspection data has confirmed that the vapor space general corrosion is not an unusually aggressive phenomena and correlates well with predicted corrosion rates for steel exposed to bulk solution. The corrosion rates are seen to decrease with time as expected. The review of the temperature data did not reveal any obvious correlations between high temperatures and the occurrences of leaks. The complex nature of temperature-humidity interaction, particularly with respect to vapor corrosion requires further understanding to infer any correlation. The review of the waste level data also did not reveal any obvious correlations.

  12. [Corrosion testing of high level radioactive waste. Final report

    SciTech Connect

    1996-06-01

    Alloys under consideration as candidates for the high level nuclear waste containers at Yucca Mountain were exposed to a range of corrosion conditions and their performance measured. The alloys tested were Incoloy 825, 70/30 Copper-Nickel, Monel 400, Hastelloy C- 22, and low carbon steel. The test conditions varied were: temperature, concentration, agitation, and crevice simulation. Only in the case of carbon steel was significant attack noted. This attack appeared to be transport limited.

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

    SciTech Connect

    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.

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

    SciTech Connect

    Peeler, David K.; Vienna, John D.; Schweiger, Michael J.; Fox, Kevin M.

    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

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

    SciTech Connect

    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.

  16. Space augmentation of military high-level waste disposal

    NASA Technical Reports Server (NTRS)

    English, T.; Lees, L.; Divita, E.

    1979-01-01

    Space disposal of selected components of military high-level waste (HLW) is considered. This disposal option offers the promise of eliminating the long-lived radionuclides in military HLW from the earth. A space mission which meets the dual requirements of long-term orbital stability and a maximum of one space shuttle launch per week over a period of 20-40 years, is a heliocentric orbit about halfway between the orbits of earth and Venus. Space disposal of high-level radioactive waste is characterized by long-term predictability and short-term uncertainties which must be reduced to acceptably low levels. For example, failure of either the Orbit Transfer Vehicle after leaving low earth orbit, or the storable propellant stage failure at perihelion would leave the nuclear waste package in an unplanned and potentially unstable orbit. Since potential earth reencounter and subsequent burn-up in the earth's atmosphere is unacceptable, a deep space rendezvous, docking, and retrieval capability must be developed.

  17. Development of a High Level Waste Tank Inspection System

    SciTech Connect

    Appel, D.K.; Loibl, M.W.; Meese, D.C.

    1995-03-21

    The Westinghouse Savannah River Technology Center was requested by it`s sister site, West Valley Nuclear Service (WVNS), to develop a remote inspection system to gather wall thickness readings of their High Level Waste Tanks. WVNS management chose to take a proactive approach to gain current information on two tanks t hat had been in service since the early 70`s. The tanks contain high level waste, are buried underground, and have only two access ports to an annular space between the tank and the secondary concrete vault. A specialized remote system was proposed to provide both a visual surveillance and ultrasonic thickness measurements of the tank walls. A magnetic wheeled crawler was the basis for the remote delivery system integrated with an off-the-shelf Ultrasonic Data Acquisition System. A development program was initiated for Savannah River Technology Center (SRTC) to design, fabricate, and test a remote system based on the Crawler. The system was completed and involved three crawlers to perform the needed tasks, an Ultrasonic Crawler, a Camera Crawler, and a Surface Prep Crawler. The crawlers were computer controlled so that their operation could be done remotely and their position on the wall could be tracked. The Ultrasonic Crawler controls were interfaced with ABB Amdata`s I-PC, Ultrasonic Data Acquisition System so that thickness mapping of the wall could be obtained. A second system was requested by Westinghouse Savannah River Company (WSRC), to perform just ultrasonic mapping on their similar Waste Storage Tanks; however, the system needed to be interfaced with the P-scan Ultrasonic Data Acquisition System. Both remote inspection systems were completed 9/94. Qualifications tests were conducted by WVNS prior to implementation on the actual tank and tank development was achieved 10/94. The second inspection system was deployed at WSRC 11/94 with success, and the system is now in continuous service inspecting the remaining high level waste tanks at WSRC.

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

    SciTech Connect

    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

  19. VITRIFICATION OF HIGH LEVEL WASTE AT THE SAVANNAH RIVER SITE

    SciTech Connect

    Fox, K.; Peeler, D.

    2009-06-17

    The objective of this study was to experimentally measure the properties and performance of a series of glasses with compositions that could represent high level waste Sludge Batch 5 (SB5) as vitrified at the Savannah River Site Defense Waste Processing Facility. These data were used to guide frit optimization efforts as the SB5 composition was finalized. Glass compositions for this study were developed by combining a series of SB5 composition projections with a group of candidate frits. The study glasses were fabricated using depleted uranium and their chemical compositions, crystalline contents and chemical durabilities were characterized. Trevorite was the only crystalline phase that was identified in a few of the study glasses after slow cooling, and is not of concern as spinels have been shown to have little impact on the durability of high level waste glasses. Chemical durability was quantified using the Product Consistency Test (PCT). All of the glasses had very acceptable durability performance. The results of this study indicate that a frit composition can be identified that will provide a processable and durable glass when combined with SB5.

  20. ATW system impact on high-level waste

    SciTech Connect

    Arthur, E.D.

    1992-12-01

    This report discusses the Accelerator Transmutation of Waste (ATW) concept which aims at destruction of key long-lived radionuclides in high-level nuclear waste (HLW), both fission products and actinides. This focus makes it different from most other transmutation concepts which concentrate primarily on actinide burning. The ATW system uses an accelerator-driven, sub-critical assembly to create an intense thermal neutron environment for radionuclide transmutation. This feature allows rapid transmutation under low-inventory system conditions, which in turn, has a direct impact on the size of chemical separations and materials handling components of the system. Inventories in ATW are factors of eight to thirty times smaller than reactor systems of equivalent thermal power. Chemical separations systems are relatively small in scale and can be optimized to achieve high decontamination factors and minimized waste streams. The low-inventory feature also directly impacts material amounts remaining in the system at its end of life. In addition to its low-inventory operation, the accelerator-driven neutron source features of ATW are key to providing a sufficient level of neutrons to allow transmutation of long-lived fission products.

  1. Advanced waste form and melter development for treatment of troublesome high-level wastes

    SciTech Connect

    Marra, James; Kim, Dong -Sang; Maio, Vincent

    2015-09-02

    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 approached 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.

  2. Deep borehole disposal of high-level radioactive waste.

    SciTech Connect

    Stein, Joshua S.; Freeze, Geoffrey A.; Brady, Patrick Vane; Swift, Peter N.; Rechard, Robert Paul; Arnold, Bill Walter; Kanney, Joseph F.; Bauer, Stephen J.

    2009-07-01

    Preliminary evaluation of deep borehole disposal of high-level radioactive waste and spent nuclear fuel indicates the potential for excellent long-term safety performance at costs competitive with mined repositories. Significant fluid flow through basement rock is prevented, in part, by low permeabilities, poorly connected transport pathways, and overburden self-sealing. Deep fluids also resist vertical movement because they are density stratified. Thermal hydrologic calculations estimate the thermal pulse from emplaced waste to be small (less than 20 C at 10 meters from the borehole, for less than a few hundred years), and to result in maximum total vertical fluid movement of {approx}100 m. Reducing conditions will sharply limit solubilities of most dose-critical radionuclides at depth, and high ionic strengths of deep fluids will prevent colloidal transport. For the bounding analysis of this report, waste is envisioned to be emplaced as fuel assemblies stacked inside drill casing that are lowered, and emplaced using off-the-shelf oilfield and geothermal drilling techniques, into the lower 1-2 km portion of a vertical borehole {approx}45 cm in diameter and 3-5 km deep, followed by borehole sealing. Deep borehole disposal of radioactive waste in the United States would require modifications to the Nuclear Waste Policy Act and to applicable regulatory standards for long-term performance set by the US Environmental Protection Agency (40 CFR part 191) and US Nuclear Regulatory Commission (10 CFR part 60). The performance analysis described here is based on the assumption that long-term standards for deep borehole disposal would be identical in the key regards to those prescribed for existing repositories (40 CFR part 197 and 10 CFR part 63).

  3. 4.5 Meter high level waste canister study

    SciTech Connect

    Calmus, R. B.

    1997-10-01

    The Tank Waste Remediation System (TWRS) Storage and Disposal Project has established the Immobilized High-Level Waste (IBLW) Storage Sub-Project to provide the capability to store Phase I and II BLW products generated by private vendors. A design/construction project, Project W-464, was established under the Sub-Project to provide the Phase I capability. Project W-464 will retrofit the Hanford Site Canister Storage Building (CSB) to accommodate the Phase I I-ILW products. Project W-464 conceptual design is currently being performed to interim store 3.0 m-long BLW stainless steel canisters with a 0.61 in diameter, DOE is considering using a 4.5 in canister of the same diameter to reduce permanent disposal costs. This study was performed to assess the impact of replacing the 3.0 in canister with the 4.5 in canister. The summary cost and schedule impacts are described.

  4. Socioeconomic studies of high-level nuclear waste disposal.

    PubMed

    White, G F; Bronzini, M S; Colglazier, E W; Dohrenwend, B; Erikson, K; Hansen, R; Kneese, A V; Moore, R; Page, E B; Rappaport, R A

    1994-11-01

    The socioeconomic investigations of possible impacts of the proposed repository for high-level nuclear waste at Yucca Mountain, Nevada, have been unprecedented in several respects. They bear on the public decision that sooner or later will be made as to where and how to dispose permanently of the waste presently at military weapons installations and that continues to accumulate at nuclear power stations. No final decision has yet been made. There is no clear precedent from other countries. The organization of state and federal studies is unique. The state studies involve more disciplines than any previous efforts. They have been carried out in parallel to federal studies and have pioneered in defining some problems and appropriate research methods. A recent annotated bibliography provides interested scientists with a compact guide to the 178 published reports, as well as to relevant journal articles and related documents. PMID:7971963

  5. Socioeconomic studies of high-level nuclear waste disposal.

    PubMed Central

    White, G F; Bronzini, M S; Colglazier, E W; Dohrenwend, B; Erikson, K; Hansen, R; Kneese, A V; Moore, R; Page, E B; Rappaport, R A

    1994-01-01

    The socioeconomic investigations of possible impacts of the proposed repository for high-level nuclear waste at Yucca Mountain, Nevada, have been unprecedented in several respects. They bear on the public decision that sooner or later will be made as to where and how to dispose permanently of the waste presently at military weapons installations and that continues to accumulate at nuclear power stations. No final decision has yet been made. There is no clear precedent from other countries. The organization of state and federal studies is unique. The state studies involve more disciplines than any previous efforts. They have been carried out in parallel to federal studies and have pioneered in defining some problems and appropriate research methods. A recent annotated bibliography provides interested scientists with a compact guide to the 178 published reports, as well as to relevant journal articles and related documents. PMID:7971963

  6. High-level waste tank farm set point document

    SciTech Connect

    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.

  7. Midwestern High-Level Radioactive Waste Transportation Project

    SciTech Connect

    Dantoin, T.S.

    1990-12-01

    For more than half a century, the Council of State Governments has served as a common ground for the states of the nation. The Council is a nonprofit, state-supported and -directed service organization that provides research and resources, identifies trends, supplies answers and creates a network for legislative, executive and judicial branch representatives. This List of Available Resources was prepared with the support of the US Department of Energy, Cooperative Agreement No. DE-FC02-89CH10402. However, any opinions, findings, conclusions, or recommendations expressed herein are those of the author(s) and do not necessarily reflect the views of DOE. The purpose of the agreement, and reports issued pursuant to it, is to identify and analyze regional issues pertaining to the transportation of high-level radioactive waste and to inform Midwestern state officials with respect to technical issues and regulatory concerns related to waste transportation.

  8. High level radioactive waste vitrification process equipment component testing

    SciTech Connect

    Siemens, D.H.; Heath, W.O.; Larson, D.E.; Craig, S.N.; Berger, D.N.; Goles, R.W.

    1985-04-01

    Remote operability and maintainability of vitrification equipment were assessed under shielded-cell conditions. The equipment tested will be applied to immobilize high-level and transuranic liquid waste slurries that resulted from plutonium production for defense weapons. Equipment tested included: a turntable for handling waste canisters under the melter; a removable discharge cone in the melter overflow section; a thermocouple jumper that extends into a shielded cell; remote instrument and electrical connectors; remote, mechanical, and heat transfer aspects of the melter glass overflow section; a reamer to clean out plugged nozzles in the melter top; a closed circuit camera to view the melter interior; and a device to retrieve samples of the glass product. A test was also conducted to evaluate liquid metals for use in a liquid metal sealing system.

  9. CEMENTITIOUS GROUT FOR CLOSING SRS HIGH LEVEL WASTE TANKS - #12315

    SciTech Connect

    Langton, C.; Burns, H.; Stefanko, D.

    2012-01-10

    In 1997, the first two United States Department of Energy (US DOE) high level waste tanks (Tanks 17-F and 20-F: Type IV, single shell tanks) were taken out of service (permanently closed) at the Savannah River Site (SRS). In 2012, the DOE plans to remove from service two additional Savannah River Site (SRS) Type IV high-level waste tanks, Tanks 18-F and 19-F. These tanks were constructed in the late 1950's and received low-heat waste and do not contain cooling coils. Operational closure of Tanks 18-F and 19-F is intended to be consistent with the applicable requirements of the Resource Conservation and Recovery Act (RCRA) and the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and will be performed in accordance with South Carolina Department of Health and Environmental Control (SCDHEC). The closure will physically stabilize two 4.92E+04 cubic meter (1.3 E+06 gallon) carbon steel tanks and isolate and stabilize any residual contaminants left in the tanks. The closure will also fill, physically stabilize and isolate ancillary equipment abandoned in the tanks. A Performance Assessment (PA) has been developed to assess the long-term fate and transport of residual contamination in the environment resulting from the operational closure of the F-Area Tank Farm (FTF) waste tanks. Next generation flowable, zero-bleed cementitious grouts were designed, tested, and specified for closing Tanks 18-F and 19-F and for filling the abandoned equipment. Fill requirements were developed for both the tank and equipment grouts. All grout formulations were required to be alkaline with a pH of 12.4 and chemically reduction potential (Eh) of -200 to -400 to stabilize selected potential contaminants of concern. This was achieved by including Portland cement and Grade 100 slag in the mixes, respectively. Ingredients and proportions of cementitious reagents were selected and adjusted, respectively, to support the mass placement strategy developed by closure

  10. CLASSIFICATION OF THE MGR DEFENSE HIGH LEVEL WASTE DISPOSAL CONTIANER

    SciTech Connect

    J.A. Ziegler

    1999-08-31

    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).

  11. Modern Alchemy: Solidifying high-level nuclear waste

    SciTech Connect

    Newton, C.C.

    1997-07-01

    The U.S. Department of Energy is putting a modern version of alchemy to work to produce an answer to a decades-old problem. It is taking place at the Savannah River Site (SRS) in Aiken, South Carolina and at the West Valley Demonstration Project (WVDP) near Buffalo, New York. At both locations, contractor Westinghouse Electric Corporation is applying technology that is turning liquid high-level radioactive waste (HLW) into a stabilized, durable glass for safer and easier management. The process is called vitrification. SRS and WVDP are now operating the nation`s first full-scale HLW vitrification plants.

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

    SciTech Connect

    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.

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

    SciTech Connect

    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/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.

  14. Tank waste remediation system high-level waste feed processability assessment report

    SciTech Connect

    Lambert, S.L.; Kim, D.S.

    1994-12-01

    This study evaluates the effect of feed composition on the performance of the high-level vitrification process. It is assumed in this study that the tank wastes are retrieved and blended by tank farms, producing 12 different blends from the single-shell tank farms, two blends of double-shell tank waste, and a separately defined all-tank blend. This blending scenario was chosen only for evaluating the impact of composition on the volume of high- level waste glass produced. Special glass compositions were formulated for each waste blend based on glass property models and the properties of similar glasses. These glasses were formulated to meet the applicable viscosity, electrical conductivity, and liquidus temperature constraints for the identified candidate melters. Candidate melters in this study include the low-temperature stirred melter, which operates at 1050{degrees}C; the reference Hanford Waste Vitrification Plant liquid-fed ceramic melter, which operates at 1150{degrees}C; and the high-temperature, joule-heated melter and the cold-crucible melter, which operate over a temperature range of 1150{degrees}C to 1400{degrees}C. In the most conservative case, it is estimated that 61,000 MT of glass will be produced if the Site`s high-level wastes are retrieved by tank farms and processed in the reference joule-heated melter. If an all-tank blend was processed under the same conditions, the reference melter would produce 21,250 MT of glass. If cross-tank blending were used, it is anticipated that $2.0 billion could be saved in repository disposal costs (based on an average disposal cost of $217,000 per canister) by blending the S, SX, B, and T Tank Farm wastes with other wastes prior to vitrification. General blending among all the tank farms is expected to produce great potential benefit.

  15. PLUTONIUM/HIGH-LEVEL VITRIFIED WASTE BDBE DOSE CALCULATION

    SciTech Connect

    J.A. Ziegler

    2000-11-20

    The purpose of this calculation is to provide a dose consequence analysis of high-level waste (HLW) consisting of plutonium immobilized in vitrified HLW to be handled at the proposed Monitored Geologic Repository at Yucca Mountain for a beyond design basis event (BDBE) under expected conditions using best estimate values for each calculation parameter. In addition to the dose calculation, a plutonium respirable particle size for dose calculation use is derived. The current concept for this waste form is plutonium disks enclosed in cans immobilized in canisters of vitrified HLW (i.e., glass). The plutonium inventory at risk used for this calculation is selected from Plutonium Immobilization Project Input for Yucca Mountain Total Systems Performance Assessment (Shaw 1999). The BDBE examined in this calculation is a nonmechanistic initiating event and the sequence of events that follow to cause a radiological release. This analysis will provide the radiological releases and dose consequences for a postulated BDBE. Results may be considered in other analyses to determine or modify the safety classification and quality assurance level of repository structures, systems, and components. This calculation uses best available technical information because the BDBE frequency is very low (i.e., less than 1.0E-6 events/year) and is not required for License Application for the Monitored Geologic Repository. The results of this calculation will not be used as part of a licensing or design basis.

  16. Crystalline plutonium hosts derived from high-level waste formulations.

    SciTech Connect

    O'Holleran, T. P.

    1998-04-24

    The Department of Energy has selected immobilization for disposal in a repository as one approach for disposing of excess plutonium (1). Materials for immobilizing weapons-grade plutonium for repository disposal must meet the ''spent fuel standard'' by providing a radiation field similar to spent fuel (2). Such a radiation field can be provided by incorporating fission products from high-level waste into the waste form. Experiments were performed to evaluate the feasibility of incorporating high-level waste (HLW) stored at the Idaho Chemical Processing Plant (ICPP) into plutonium dispositioning materials to meet the spent fuel standard. A variety of materials and preparation techniques were evaluated based on prior experience developing waste forms for immobilizing HLW. These included crystalline ceramic compositions prepared by conventional sintering and hot isostatic pressing (HIP), and glass formulations prepared by conventional melting. Because plutonium solubility in silicate melts is limited, glass formulations were intentionally devitrified to partition plutonium into crystalline host phases, thereby allowing increased overall plutonium loading. Samarium, added as a representative rare earth neutron absorber, also tended to partition into the plutonium host phases. Because the crystalline plutonium host phases are chemically more inert, the plutonium is more effectively isolated from the environment, and its attractiveness for proliferation is reduced. In the initial phase of evaluating each material and preparation method, cerium was used as a surrogate for plutonium. For promising materials, additional preparation experiments were performed using plutonium to verify the behavior of cerium as a surrogate. These experiments demonstrated that cerium performed well as a surrogate for plutonium. For the most part, cerium and plutonium partitioned onto the same crystalline phases, and no anomalous changes in oxidation state were observed. The only observed

  17. Ultrafilter Conditions for High Level Waste Sludge Processing

    SciTech Connect

    Geeting, John GH; Hallen, Richard T.; Peterson, Reid A.

    2006-08-28

    An evaluation of the optimal filtration conditions was performed based on test data obtained from filtration of a High Level Waste Sludge sample from the Hanford tank farms. This evaluation was performed using the anticipated configuration for the Waste Treatment Plant at the Hanford site. Testing was performed to identify the optimal pressure drop and cross flow velocity for filtration at both high and low solids loading. However, this analysis indicates that the actual filtration rate achieved is relatively insensitive to these conditions under anticipated operating conditions. The maximum filter flux was obtained by adjusting the system control valve pressure from 400 to 650 kPa while the filter feed concentration increased from 5 to 20 wt%. However, operating the system with a constant control valve pressure drop of 500 kPa resulted in a less than 1% reduction in the average filter flux. Also note that allowing the control valve pressure to swing as much as +/- 20% resulted in less than a 5% decrease in filter flux.

  18. Waste package performance evaluations for the proposed high-level nuclear waste repository at Yucca Mountain.

    PubMed

    Mon, Kevin G; Bullard, Bryan E; Mehta, Sunil; Lee, Joon H

    2004-04-01

    The evaluation studies of the proposed repository for long-term storage of spent nuclear fuel and high-level nuclear waste at Yucca Mountain, Nevada, are underway. Fulfillment of the requirements for limiting dose to the public, which includes containment of the radioactive waste emplaced in the proposed repository and subsequent slow release of radionuclides from the Engineered Barrier System (EBS) into the geosphere, will rely on a robust waste container design, among other EBS components. Part of the evaluation process involves sensitivity studies aimed at elucidating which model parameters contribute most to the waste package and overlying drip shield degradation characteristics. The model parameters identified for this study include (1) general corrosion rate parameters and (2) stress corrosion cracking (SCC) parameters. Temperature dependence and parameter uncertainty are evaluated for the general corrosion rate model parameters while for the SCC model parameters, uncertainty treatment of stress intensity factor, crack initiation threshold, and manufacturing flaw orientations are evaluated. Based on these evaluations new uncertainty distributions are generated and recommended for future analyses. Also, early waste package failures due to improper heat treatment were added to the waste package degradation model. The results of these investigations indicate that the waste package failure profiles are governed by the manufacturing flaw orientation model parameters. PMID:15078313

  19. Technical considerations for evaluating substantially complete containment of high-level waste within the waste package

    SciTech Connect

    Manaktala, H.K. . Center for Nuclear Waste Regulatory Analyses); Interrante, C.G. . Div. of High-Level Waste Management)

    1990-12-01

    This report deals with technical information that is considered essential for demonstrating the ability of the high-level radioactive waste package to provide substantially complete containment'' of its contents (vitrified waste form or spent light-water reactor fuel) for a period of 300 to 1000 years in a geological repository environment. The discussion is centered around technical considerations of the repository environment, materials and fabrication processes for the waste package components, various degradation modes of the materials of construction of the waste packages, and inspection and monitoring of the waste package during the preclosure and retrievability period, which could begin up to 50 years after initiation of waste emplacement. The emphasis in this report is on metallic materials. However, brief references have been made to other materials such as ceramics, graphite, bonded ceramic-metal systems, and other types of composites. The content of this report was presented to an external peer review panel of nine members at a workshop held at the Center for Nuclear Waste Regulatory Analyses (CNWRA), Southwest Research Institute, San Antonio, Texas, April 2--4, 1990. The recommendations of the peer review panel have been incorporated in this report. There are two companion reports; the second report in the series provides state-of-the-art techniques for uncertainty evaluations. 97 refs., 1 fig.

  20. JET MIXING ANALYSIS FOR SRS HIGH-LEVEL WASTE RECOVERY

    SciTech Connect

    Lee, S.

    2011-07-05

    The process of recovering the waste in storage tanks at the Savannah River Site (SRS) typically requires mixing the contents of the tank to ensure uniformity of the discharge stream. Mixing is accomplished with one to four slurry pumps located within the tank liquid. The slurry pump may be fixed in position or they may rotate depending on the specific mixing requirements. The high-level waste in Tank 48 contains insoluble solids in the form of potassium tetraphenyl borate compounds (KTPB), monosodium titanate (MST), and sludge. Tank 48 is equipped with 4 slurry pumps, which are intended to suspend the insoluble solids prior to transfer of the waste to the Fluidized Bed Steam Reformer (FBSR) process. The FBSR process is being designed for a normal feed of 3.05 wt% insoluble solids. A chemical characterization study has shown the insoluble solids concentration is approximately 3.05 wt% when well-mixed. The project is requesting a Computational Fluid Dynamics (CFD) mixing study from SRNL to determine the solids behavior with 2, 3, and 4 slurry pumps in operation and an estimate of the insoluble solids concentration at the suction of the transfer pump to the FBSR process. The impact of cooling coils is not considered in the current work. The work consists of two principal objectives by taking a CFD approach: (1) To estimate insoluble solids concentration transferred from Tank 48 to the Waste Feed Tank in the FBSR process and (2) To assess the impact of different combinations of four slurry pumps on insoluble solids suspension and mixing in Tank 48. For this work, several different combinations of a maximum of four pumps are considered to determine the resulting flow patterns and local flow velocities which are thought to be associated with sludge particle mixing. Two different elevations of pump nozzles are used for an assessment of the flow patterns on the tank mixing. Pump design and operating parameters used for the analysis are summarized in Table 1. The baseline

  1. PLUTONIUM/HIGH-LEVEL VITRIFIED WASTE BDBE DOSE CALCULATION

    SciTech Connect

    D.C. Richardson

    2003-03-19

    In accordance with the Nuclear Waste Policy Amendments Act of 1987, Yucca Mountain was designated as the site to be investigated as a potential repository for the disposal of high-level radioactive waste. The Yucca Mountain site is an undeveloped area located on the southwestern edge of the Nevada Test Site (NTS), about 100 miles northwest of Las Vegas. The site currently lacks rail service or an existing right-of-way. If the Yucca Mountain site is found suitable for the repository, rail service is desirable to the Office of Civilian Waste Management (OCRWM) Program because of the potential of rail transportation to reduce costs and to reduce the number of shipments relative to highway transportation. A Preliminary Rail Access Study evaluated 13 potential rail spur options. Alternative routes within the major options were also developed. Each of these options was then evaluated for potential land use conflicts and access to regional rail carriers. Three potential routes having few land use conflicts and having access to regional carriers were recommended for further investigation. Figure 1-1 shows these three routes. The Jean route is estimated to be about 120 miles long, the Carlin route to be about 365 miles long, and Caliente route to be about 365 miles long. The remaining ten routes continue to be monitored and should any of the present conflicts change, a re-evaluation of that route will be made. Complete details of the evaluation of the 13 routes can be found in the previous study. The DOE has not identified any preferred route and recognizes that the transportation issues need a full and open treatment under the National Environmental Policy Act. The issue of transportation will be included in public hearings to support development of the Environmental Impact Statement (EIS) proceedings for either the Monitored Retrievable Storage Facility or the Yucca Mountain Project or both.

  2. Stability of High-Level Radioactive Waste Forms

    SciTech Connect

    Besmann, T.M.

    2001-06-22

    High-level waste (HLW) glass compositions, processing schemes, limits on waste content, and corrosion/dissolution release models are dependent on an accurate knowledge of melting temperatures and thermochemical values. Unfortunately, existing models for predicting these temperatures are empirically-based, depending on extrapolations of experimental information. In addition, present models of leaching behavior of glass waste forms use simplistic assumptions or experimentally measured values obtained under non-realistic conditions. There is thus a critical need for both more accurate and more widely applicable models for HLW glass behavior, which this project addressed. Significant progress was made in this project on modeling HLW glass. Borosilicate glass was accurately represented along with the additional important components that contain iron, lithium, potassium, magnesium, and calcium. The formation of crystalline inclusions in the glass, an issue in Hanford HLW formulations, was modeled and shown to be predictive. Thus the results of this work have already demonstrated practical benefits with the ability to map compositional regions where crystalline material forms, and therefore avoid that detrimental effect. With regard to a fundamental understanding, added insights on the behavior of the components of glass have been obtained, including the potential formation of molecular clusters. The EMSP project had very significant effects beyond the confines of Environmental Management. The models developed for glass have been used to solve a very costly problem in the corrosion of refractories for glass production. The effort resulted in another laboratory, Sandia National Laboratories-Livermore, to become conversant in the techniques and to apply those through a DOE Office of Industrial Technologies project joint with PPG Industries. The glass industry as a whole is now cognizant of these capabilities, and there is a Glass Manufacturer's Research Institute proposal

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

    SciTech Connect

    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.

  4. Calculates Neutron Production in Canisters of High-level Waste

    Energy Science and Technology Software Center (ESTSC)

    1993-01-15

    ALPHN calculates the (alpha,n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the (alpha,n) neutron production of each actinide in neutrons per second and the total for the canister. The (alpha,n) neutron production rates are source terms only; that is, they are production rates within the glass andmore » do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister.« less

  5. Strontium and Actinide Separations from High Level Nuclear Waste Solutions using Monosodium Titanate - Actual Waste Testing

    SciTech Connect

    Peters, T.B.; Barnes, M.J.; Hobbs,D.T.; Walker, D.D.; Fondeur, F.F.; Norato, M.A.; Pulmano, R.L.; Fink, S.D.

    2005-11-01

    Pretreatment processes at the Savannah River Site will separate {sup 90}Sr, alpha-emitting and radionuclides (i.e., actinides) and {sup 137}Cs prior to disposal of the high-level nuclear waste. Separation of {sup 90}Sr and alpha-emitting radionuclides occurs by ion exchange/adsorption using an inorganic material, monosodium titanate (MST). Previously reported testing with simulants indicates that the MST exhibits high selectivity for strontium and actinides in high ionic strength and strongly alkaline salt solutions. This paper provides a summary of data acquired to measure the performance of MST to remove strontium and actinides from actual waste solutions. These tests evaluated the effects of ionic strength, mixing, elevated alpha activities, and multiple contacts of the waste with MST. Tests also provided confirmation that MST performs well at much larger laboratory scales (300-700 times larger) and exhibits little affinity for desorption of strontium and plutonium during washing.

  6. THERMAL ANALYSIS OF GEOLOGIC HIGH-LEVEL RADIOACTIVE WASTE PACKAGES

    SciTech Connect

    Hensel, S.; Lee, S.

    2010-04-20

    The engineering design of disposal of the high level waste (HLW) packages in a geologic repository requires a thermal analysis to provide the temperature history of the packages. Calculated temperatures are used to demonstrate compliance with criteria for waste acceptance into the geologic disposal gallery system and as input to assess the transient thermal characteristics of the vitrified HLW Package. The objective of the work was to evaluate the thermal performance of the supercontainer containing the vitrified HLW in a non-backfilled and unventilated underground disposal gallery. In order to achieve the objective, transient computational models for a geologic vitrified HLW package were developed by using a computational fluid dynamics method, and calculations for the HLW disposal gallery of the current Belgian geological repository reference design were performed. An initial two-dimensional model was used to conduct some parametric sensitivity studies to better understand the geologic system's thermal response. The effect of heat decay, number of co-disposed supercontainers, domain size, humidity, thermal conductivity and thermal emissivity were studied. Later, a more accurate three-dimensional model was developed by considering the conduction-convection cooling mechanism coupled with radiation, and the effect of the number of supercontainers (3, 4 and 8) was studied in more detail, as well as a bounding case with zero heat flux at both ends. The modeling methodology and results of the sensitivity studies will be presented.

  7. High Level Waste Feed Certification in Hanford Double Shell Tanks

    SciTech Connect

    Thien, Micheal G.; Wells, Beric E.; Adamson, Duane J.

    2010-03-01

    The ability to effectively mix, sample, certify, and deliver consistent batches of High Level Waste (HLW) feed from the Hanford Double Shell Tanks (DST) 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. DOE’s River Protection Project (RPP) mission modeling and WTP facility modeling assume that individual 3785 cubic meter (1 million gallon) HLW feed tanks are homogenously mixed, representatively sampled, and consistently delivered to the WTP. It has been demonstrated that homogenous mixing of HLW sludge in Hanford DSTs is not likely achievable with the baseline design thereby causing representative sampling and consistent feed delivery to be more difficult. Inconsistent feed to the WTP could cause additional batch to batch operational adjustments that reduces operating efficiency and has the potential to increase the overall mission length. The Hanford mixing and sampling demonstration program will identify DST mixing performance capability, will evaluate representative sampling techniques, and will estimate feed batch consistency. An evaluation of demonstration program results will identify potential mission improvement considerations that will help ensure successful mission completion. This paper will discuss the history, progress, and future activities that will define and mitigate the mission risk.

  8. MECHANISMS AND KINETICS OF ORGANIC AGING IN HIGH-LEVEL NUCLEAR WASTES

    EPA Science Inventory

    Highly radioactive wastes stored at Hanford and Savannah River DOE sites have unresolved questions relating to safety of the stored waste, as well as needs for safe, effective, and efficient waste processing to minimize the volume of high-level waste (HLW) streams for disposal. ...

  9. Removal of cesium from a high-level calcined waste by high temperature volatilization

    SciTech Connect

    Del Debbio, J.A.

    1994-11-01

    Pyrochemical methods are being evaluated for the separation of actinides and fission products from inert material in high-level waste calcine. Separation processes have the potential of reducing waste disposal costs by reducing the volume of high-level waste requiring final disposal in a repository. Tests were conducted to evaluate high temperature volatilization for removing {sup 137}Cs from four types of calcines. The results for pilot plant calcines indicate greater than 99% cesium removal for alumina and fluorinel/sodium calcines heated at 1000{degree}C and 99% removal for zirconia calcine heated at 1100{degree}C. Tests with actual calcine generated at the New Waste Calcining Facility (NWCF) from a blend of aluminium, sodium and zirconium/fluorinel wastes resulted in 96% cesium removal at 1000{degree}C and greater than 99% removal at 1170{degree}C.

  10. Low-level waste disposal in highly populated areas

    SciTech Connect

    Kowalski, E.; McCombie, C.; Issler, H.

    1989-11-01

    Nuclear-generated electricity supplies almost 40% of the demand in Switzerland (the rest being hydro-power). Allowing for a certain reserve and assuming an operational life-time of 40 years for each reactor, and taking into account wastes from decommissioning and from medicine, industry and research, the total amount of low-level radioactive waste to be disposed of is about 175,000 m{sup 3}. Since there are no unpopulated areas in Switzerland, and since Swiss Federal Law specifies that the safety of disposal may not depend upon supervision of the repository, no shallow-land burial has been foreseen, even for short-lived low-level waste. Instead, geological disposal in a mined cavern system with access through a horizontal tunnel was selected as the best way of meeting the requirements and ensuring the necessary public acceptance.

  11. Characteristics Data Base: Programmer's guide to the High-Level Waste Data Base

    SciTech Connect

    Jones, K.E. ); Salmon, R. )

    1990-08-01

    The High-Level Waste Data Base is a menu-driven PC data base developed as part of OCRWM's technical data base on the characteristics of potential repository wastes, which also includes spent fuel and other materials. This programmer's guide completes the documentation for the High-Level Waste Data Base, the user's guide having been published previously. 3 figs.

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

    SciTech Connect

    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{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.

  13. Cementitious Grout for Closing SRS High Level Waste Tanks - 12315

    SciTech Connect

    Langton, C.A.; Stefanko, D.B.; Burns, H.H.; Waymer, J.; Mhyre, W.B.; Herbert, J.E.; Jolly, J.C. Jr.

    2012-07-01

    In 1997, the first two United States Department of Energy (US DOE) high level waste tanks (Tanks 17-F and 20-F: Type IV, single shell tanks) were taken out of service (permanently closed) at the Savannah River Site (SRS). In 2012, the DOE plans to remove from service two additional Savannah River Site (SRS) Type IV high-level waste tanks, Tanks 18-F and 19-F. These tanks were constructed in the late 1950's and received low-heat waste and do not contain cooling coils. Operational closure of Tanks 18-F and 19-F is intended to be consistent with the applicable requirements of the Resource Conservation and Recovery Act (RCRA) and the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and will be performed in accordance with South Carolina Department of Health and Environmental Control (SCDHEC). The closure will physically stabilize two 4.92E+04 cubic meter (1.3 E+06 gallon) carbon steel tanks and isolate and stabilize any residual contaminants left in the tanks. Ancillary equipment abandoned in the tanks will also be filled to the extent practical. A Performance Assessment (PA) has been developed to assess the long-term fate and transport of residual contamination in the environment resulting from the operational closure of the F-Area Tank Farm (FTF) waste tanks. Next generation flowable, zero-bleed cementitious grouts were designed, tested, and specified for closing Tanks 18-F and 19-F and for filling the abandoned equipment. Fill requirements were developed for both the tank and equipment grouts. All grout formulations were required to be alkaline with a pH of 12.4 and to be chemically reducing with a reduction potential (Eh) of -200 to -400. Grouts with this chemistry stabilize potential contaminants of concern. This was achieved by including Portland cement and Grade 100 slag in the mixes, respectively. Ingredients and proportions of cementitious reagents were selected and adjusted to support the mass placement strategy developed by

  14. Production of a High-Level Waste Glass from Hanford Waste Samples

    SciTech Connect

    Crawford, C.L.; Farrara, D.M.; Ha, B.C.; Bibler, N.E.

    1998-09-01

    The HLW glass was produced from a HLW sludge slurry (Envelope D Waste), eluate waste streams containing high levels of Cs-137 and Tc-99, solids containing both Sr-90 and transuranics (TRU), and glass-forming chemicals. The eluates and Sr-90/TRU solids were obtained from ion-exchange and precipitation pretreatments, respectively, of other Hanford supernate samples (Envelopes A, B and C Waste). The glass was vitrified by mixing the different waste streams with glass-forming chemicals in platinum/gold crucibles and heating the mixture to 1150 degree C. Resulting glass analyses indicated that the HLW glass waste form composition was close to the target composition. The targeted waste loading of Envelope D sludge solids in the HLW glass was 30.7 wt percent, exclusive of Na and Si oxides. Condensate samples from the off-gas condenser and off-gas dry-ice trap indicated that very little of the radionuclides were volatilized during vitrification. Microstructure analysis of the HLW glass using Scanning Electron Microscopy (SEM) and Energy Dispersive X-Ray Analysis (EDAX) showed what appeared to be iron spinel in the HLW glass. Further X-Ray Diffraction (XRD) analysis confirmed the presence of nickel spinel trevorite (NiFe2O4). These crystals did not degrade the leaching characteristics of the glass. The HLW glass waste form passed leach tests that included a standard 90 degree C Product Consistency Test (PCT) and a modified version of the United States Environmental Protection Agency Toxicity Characteristic Leaching Procedure (TCLP).

  15. Why consider subseabed disposal of high-level nuclear waste

    SciTech Connect

    Heath, G. R.; Hollister, C. D.; Anderson, D. R.; Leinen, M.

    1980-01-01

    Large areas of the deep seabed warrant assessment as potential disposal sites for high-level radioactive waste because: (1) they are far from seismically and tectonically active lithospheric plate boundaries; (2) they are far from active or young volcanos; (3) they contain thick layers of very uniform fine-grained clays; (4) they are devoid of natural resources likely to be exploited in the forseeable future; (5) the geologic and oceanographic processes governing the deposition of sediments in such areas are well understood, and are remarkably insensitive to past oceanographic and climatic changes; and (6) sedmentary records of tens of millions of years of slow, uninterrupted deposition of fine grained clay support predictions of the future stability of such sites. Data accumulated to date on the permeability, ion-retardation properties, and mechanical strength of pelagic clay sediments indicate that they can act as a primary barrier to the escape of buried nuclides. Work in progress should determine within the current decade whether subseabed disposal is environmentally acceptable and technically feasible, as well as address the legal, political and social issues raised by this new concept.

  16. Development of polyphase ceramics for the immobilization of high-level Defense nuclear waste

    SciTech Connect

    Morgan, P.E.D.; Harker, A.B.; Clarke, D.R.; Flintoff, J.J.; Shaw, T.M.

    1983-02-25

    The report contains two major sections: Section I - An Improved Polyphase Ceramic for High-Level Defense Nucleation Waste reports the work conducted on titanium-silica based ceramics for immobilizing Savannah River Plant waste. Section II - Formulation and Processing of Alumina Based Ceramic Nuclear Waste Forms describes the work conducted on developing a generic alumina and alumina-silica based ceramic waste form capable of immobilizing any nuclear waste with a high aluminum content. Such wastes include the Savannah River Plant wastes, Hanford neutralized purex wastes, and Hanford N-Reactor acid wastes. The design approach and process technology in the two reports demonstrate how the generic high waste loaded ceramic form can be applied to a broad range of nuclear waste compositions. The individual sections are abstracted and indexed separately.

  17. High level waste interim storge architecture selection - decision report

    SciTech Connect

    Calmus, R.B.

    1996-09-27

    The U.S. Department of Energy (DOE) has embarked upon a course to acquire Hanford Site tank waste treatment and immobilization services using privatized facilities (RL 1996a). This plan contains a two-phased approach. Phase I is a proof-of-principle/connnercial demonstration- scale effort and Phase II is a fiill-scale production effort. In accordance with the planned approach, interim storage and disposal of various products from privatized facilities are to be DOE fumished. The high-level waste (BLW) interim storage options, or alternative architectures, were identified and evaluated to provide the framework from which to select the most viable method of Phase I BLW interim storage (Calmus 1996). This evaluation, hereafter referred to as the Alternative Architecture Evaluation, was performed to established performance and risk criteria (technical merit, cost, schedule, etc.). Based on evaluation results, preliminary architectures and path forward reconunendations were provided for consideration in the architecture decision- maldng process. The decision-making process used for selection of a Phase I solidified BLW interim storage architecture was conducted in accordance with an approved Decision Plan (see the attachment). This decision process was based on TSEP-07,Decision Management Procedure (WHC 1995). The established decision process entailed a Decision Board, consisting of Westinghouse Hanford Company (VY`HC) management staff, and included appointment of a VTHC Decision Maker. The Alternative Architecture Evaluation results and preliminary recommendations were presented to the Decision Board members for their consideration in the decision-making process. The Alternative Architecture Evaluation was prepared and issued before issuance of @C-IP- 123 1, Alternatives Generation and Analysis Procedure (WI-IC 1996a), but was deemed by the Board to fully meet the intent of WHC-IP-1231. The Decision Board members concurred with the bulk of the Alternative Architecture

  18. Demonstration of Caustic-Side Solvent Extraction with Savannah River Site High Level Waste

    SciTech Connect

    Walker, D.D.

    2001-08-27

    Researchers successfully demonstrated the chemistry and process equipment of the Caustic-Side Solvent Extraction (CSSX) flowsheet for the decontamination of high level waste using a 33-stage, 2-cm centrifugal contactor apparatus at the Savannah River Technology Center. This represents the first CSSX process demonstration using Savannah River Site (SRS) high level waste. Three tests lasting 6, 12, and 48 hours processed simulated average SRS waste, simulated Tank 37H/44F composite waste, and Tank 37H/44F high level waste, respectively.

  19. Mercury Reduction and Removal from High Level Waste at the Defense Waste Processing Facility - 12511

    SciTech Connect

    Behrouzi, Aria; Zamecnik, Jack

    2012-07-01

    The Defense Waste Processing Facility processes legacy nuclear waste generated at the Savannah River Site during production of enriched uranium and plutonium required by the Cold War. The nuclear waste is first treated via a complex sequence of controlled chemical reactions and then vitrified into a borosilicate glass form and poured into stainless steel canisters. Converting the nuclear waste into borosilicate glass is a safe, effective way to reduce the volume of the waste and stabilize the radionuclides. One of the constituents in the nuclear waste is mercury, which is present because it served as a catalyst in the dissolution of uranium-aluminum alloy fuel rods. At high temperatures mercury is corrosive to off-gas equipment, this poses a major challenge to the overall vitrification process in separating mercury from the waste stream prior to feeding the high temperature melter. Mercury is currently removed during the chemical process via formic acid reduction followed by steam stripping, which allows elemental mercury to be evaporated with the water vapor generated during boiling. The vapors are then condensed and sent to a hold tank where mercury coalesces and is recovered in the tank's sump via gravity settling. Next, mercury is transferred from the tank sump to a purification cell where it is washed with water and nitric acid and removed from the facility. Throughout the chemical processing cell, compounds of mercury exist in the sludge, condensate, and off-gas; all of which present unique challenges. Mercury removal from sludge waste being fed to the DWPF melter is required to avoid exhausting it to the environment or any negative impacts to the Melter Off-Gas system. The mercury concentration must be reduced to a level of 0.8 wt% or less before being introduced to the melter. Even though this is being successfully accomplished, the material balances accounting for incoming and collected mercury are not equal. In addition, mercury has not been effectively

  20. Reevaluation Of Vitrified High-Level Waste Form Criteria For Potential Cost Savings At The Defense Waste Processing Facility

    SciTech Connect

    Ray, J. W.; Marra, S. L.; Herman, C. C.

    2013-01-09

    At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive high level waste (HLW) sludge into a durable borosilicate glass since 1996. Currently the DWPF has poured over 3,500 canisters, all of which are compliant with the U. S. Department of Energy's (DOE) Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS) and therefore ready to be shipped to a federal geologic repository for permanent disposal. Due to DOE petitioning to withdraw the Yucca Mountain License Application (LA) from the Nuclear Regulatory Commission (NRC) in 2010 and thus no clear disposal path for SRS canistered waste forms, there are opportunities for cost savings with future canister production at DWPF and other DOE producer sites by reevaluating high-level waste form requirements and compliance strategies and reducing/eliminating those that will not negatively impact the quality of the canistered waste form.

  1. Reevaluation of Vitrified High-Level Waste Form Criteria for Potential Cost Savings at the Defense Waste Processing Facility - 13598

    SciTech Connect

    Ray, J.W.; Marra, S.L.; Herman, C.C.

    2013-07-01

    At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive high level waste (HLW) sludge into a durable borosilicate glass since 1996. Currently the DWPF has poured over 3,500 canisters, all of which are compliant with the U. S. Department of Energy's (DOE) Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS) and therefore ready to be shipped to a federal geologic repository for permanent disposal. Due to DOE petitioning to withdraw the Yucca Mountain License Application (LA) from the Nuclear Regulatory Commission (NRC) in 2010 and thus no clear disposal path for SRS canistered waste forms, there are opportunities for cost savings with future canister production at DWPF and other DOE producer sites by reevaluating high-level waste form requirements and compliance strategies and reducing/eliminating those that will not negatively impact the quality of the canistered waste form. (authors)

  2. Department of Energy pretreatment of high-level and low-level wastes

    SciTech Connect

    McGinnis, C.P.; Hunt, R.D.

    1995-12-31

    The remediation of the 1 {times} 10{sup 8} gal of highly radioactive waste in the underground storage tanks (USTs) at five US Department of Energy (DOE) sites is one of DOE`s greatest challenges. Therefore, the DOE Office of Environmental Management has created the Tank Focus Area (TFA) to manage an integrated technology development program that results in the safe and efficient remediation of UST waste. The TFA has divided its efforts into five areas, which are safety, characterization, retrieval/closure, pretreatment, and immobilization. All DOE pretreatment activities are integrated by the Pretreatment Technical Integration Manager of the TFA. For FY 1996, the 14 pretreatment tasks are divided into 3 systems: supernate separations, sludge treatment, and solid/liquid separation. The plans and recent results of these TFA tasks, which include two 25,000-gal demonstrations and two former TFA tasks on Cs removal, are presented. The pretreatment goals are to minimize the volume of high-level waste and the radioactivity in low-level waste.

  3. High-Level Waste Systems Plan. Revision 7

    SciTech Connect

    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.

  4. LIQUIDUS TEMPERATURE AND PRIMARY CRYSTALLIZATION PHASES IN HIGH-ZIRCONIA HIGH-LEVEL WASTE BOROSILICATE GLASSES

    EPA Science Inventory

    Liquidus temperature (TL) studies of high-Zr high-level waste (HLW) borosilicate glasseshave identified three primary phases: baddelyite (ZrO2), zircon (ZrSiO4), and alkali-zirconiumsilicates, such as parakeldyshite (Na2ZrSi2O7). Using published TL data for HLW glasses withthe...

  5. Assessment of high-level waste form conformance with proposed regulatory and repository criteria

    SciTech Connect

    Gordon, D E; Gray, P L; Jennings, A S; Permar, P H

    1982-04-01

    Federal regulatory criteria for geologic disposal of high-level waste are under development. Also, interim performance specifications for high-level waste forms in geologic isolation are being developed within the Federal program responsible for repository selection and operation. Two high-level waste forms, borosilicate glass and crystalline ceramic, have been selected as candidate immobilization forms for the Defense Waste Processing Facility (DWPF) which is to immobilize high-level wastes at the Savannah River Plant (SRP). An assessment of how these two waste forms conform with the proposed regulatory criteria and repository specifications was performed. Both forms were determined to be in conformance with postulated rules for radionuclide releases and radiation exposures throughout the entire waste disposal system, as well as with proposed repository operation requirements.

  6. Development of Crystal-Tolerant High-Level Waste Glasses

    SciTech Connect

    Matyas, Josef; Vienna, John D.; Schaible, Micah J.; Rodriguez, Carmen P.; Crum, Jarrod V.; Arrigoni, Alyssa L.; Tate, Rachel M.

    2010-12-17

    Twenty five glasses were formulated. They were batched from HLW AZ-101 simulant or raw chemicals and melted and tested with a series of tests to elucidate the effect of spinel-forming components (Ni, Fe, Cr, Mn, and Zn), Al, and noble metals (Rh2O3 and RuO2) on the accumulation rate of spinel crystals in the glass discharge riser of the high-level waste (HLW) melter. In addition, the processing properties of glasses, such as the viscosity and TL, were measured as a function of temperature and composition. Furthermore, the settling of spinel crystals in transparent low-viscosity fluids was studied at room temperature to access the shape factor and hindered settling coefficient of spinel crystals in the Stokes equation. The experimental results suggest that Ni is the most troublesome component of all the studied spinel-forming components producing settling layers of up to 10.5 mm in just 20 days in Ni-rich glasses if noble metals or a higher concentration of Fe was not introduced in the glass. The layer of this thickness can potentially plug the bottom of the riser, preventing glass from being discharged from the melter. The noble metals, Fe, and Al were the components that significantly slowed down or stopped the accumulation of spinel at the bottom. Particles of Rh2O3 and RuO2, hematite and nepheline, acted as nucleation sites significantly increasing the number of crystals and therefore decreasing the average crystal size. The settling rate of ≤10-μm crystal size around the settling velocity of crystals was too low to produce thick layers. The experimental data for the thickness of settled layers in the glasses prepared from AZ-101 simulant were used to build a linear empirical model that can predict crystal accumulation in the riser of the melter as a function of concentration of spinel-forming components in glass. The developed model predicts the thicknesses of accumulated layers quite well, R2 = 0.985, and can be become an efficient tool for the formulation

  7. Impact testing of simulated high-level waste glass canisters

    SciTech Connect

    Peterson, M.E.; Alzheimer, J.M.; Slate, S.C.

    1985-01-01

    Three Savannah River Laboratory reference high-level waste canisters were subjected to impact tests at the Pacific Northwest Laboratory in Richland, Washington, in June 1983. The purpose of the test was to determine the integrity of the canister, nozzle, and final closure weld and to assess the effects of impacts on the glass. Two of the canisters were fabricated from 304L stainless steel and the third canister from titanium. The titanium canister was subjected to two drops. The first drop was vertical from 9.14 m onto an unyielding surface with the bottom corner of the canister receiving the impact. No failure occurred during this drop. The second drop was vertical from 9.14 m onto an unyielding surface with the corner of the fill nozzle receiving the impact. A large breach in the canister occurred in the region where the fill nozzle joins the dished head. The first stainless steel canister was dropped with the corner of the fill nozzle receiving the impact. The canister showed significant strain with no rupturing in the region where the fill nozzle joins the dished head. The second canister was dropped with the bottom corner receiving the impact and also, dropped horizontally onto an unyielding vertical solid steel cylinder in a puncture test. The bottom drop did not damage the weld and the puncture test did not rupture the canister body. The glass particles in the damaged zone of these canisters were sampled and analyzed for particle size. A comparison was made with control canister in which no impact had occurred. The particle size distribution for the control canisters and the zones of damaged glass were determined down to 1.5 ..mu..m. The quantity of glass fines, smaller than 10 ..mu..m, which must be determined for transportation safety studies, was found to be the largest in the bottom-damaged zone. The total amount of fines smaller than 10 ..mu..m after impact was less than 0.01 wt % of the total amount of glass in the canister.

  8. 3-D MAPPING TECHNOLOGIES FOR HIGH LEVEL WASTE TANKS

    SciTech Connect

    Marzolf, A.; Folsom, M.

    2010-08-31

    time-of-flight data (3D image) collected with a single laser pulse, high frame rates, direct calculation of range, blur-free images without motion distortion, no need for precision scanning mechanisms, ability to combine 3D flash LIDAR with 2D cameras for 2D texture over 3D depth, and no moving parts. The major disadvantage of the 3D flash LIDAR camera is the cost of approximately $150,000, not including the software development time and repackaging of the camera for deployment in the waste tanks.

  9. Special Analysis for Disposal of High-Concentration I-129 Waste in the Intermediate-Level Vaults at the E-Area Low-Level Waste Facility

    SciTech Connect

    Collard, L.B.

    2000-09-26

    This revision was prepared to address comments from DOE-SR that arose following publication of revision 0. This Special Analysis (SA) addresses disposal of wastes with high concentrations of I-129 in the Intermediate-Level (IL) Vaults at the operating, low-level radioactive waste disposal facility (the E-Area Low-Level Waste Facility or LLWF) on the Savannah River Site (SRS). This SA provides limits for disposal in the IL Vaults of high-concentration I-129 wastes, including activated carbon beds from the Effluent Treatment Facility (ETF), based on their measured, waste-specific Kds.

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

    SciTech Connect

    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.

  11. PLUTONIUM/HIGH LEVEL VITRIFIED WASTE - DBE OFFSITE DOSE CALCULATION

    SciTech Connect

    S. O. Bader

    1999-09-20

    The purpose of this calculation is to provide a bounding dose consequence analysis of the immobilized plutonium (can-in-canister) waste form to be handled at the Monitored Geologic Repository (MGR) at Yucca Mountain. The current concept for the Plutonium Can-in-Canister waste form is provided in Attachment III. A typical design basis event (DBE) defines a scenario that generally includes an initiating event and the sequences of events that follow. This analysis will provide (1) radiological releases and dose consequences for a postulated, bounding DBE and (2) design-related assumptions on which the calculated dose consequences are based. This analysis is part of the safety design basis for the repository. Results will be used in other analyses to determine or modify the safety classification and quality assurance level of repository structures, systems, and components (SSCs). The Quality Assurance (QA) program applies to this calculation. The work reported in this document is part of the analysis of MGR DBEs and is performed using AP-3.12Q, Calculations. The work done for this analysis was evaluated according to QAP-2-0, Control of Activities. This evaluation determined that such activities are subject to DOE/RW/0333PY Quality Assurance Requirements and Description (DOE 1998), requirements. This calculation is quality affecting because the results may be used to support analyses of repository SSCs per QAP-2-3, Classification of Permanent Items.

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

    NASA Astrophysics Data System (ADS)

    1980-01-01

    Information concerning the condition of the high-level waste tanks at the Western New York State Nuclear Service center near West Valley, New York is presented. This information is 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. Immobilized high-level waste interim storage alternatives generation and analysis and decision report

    SciTech Connect

    CALMUS, R.B.

    1999-05-18

    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.

  14. A proposed classification system for high-level and other radioactive wastes

    SciTech Connect

    Kocher, D. C.; Croff, A. G.

    1987-06-01

    This report presents a proposal for quantitative and generally applicable risk-based definitions of high-level and other radioactive wastes. On the basis of historical descriptions and definitions of high-level waste (HLW), in which HLW has been defined in terms of its source as waste from reprocessing of spent nuclear fuel, we propose a more general definition based on the concept that HLW has two distinct attributes: HLW is (1) highly radioactive and (2) requires permanent isolation. This concept leads to a two-dimensional waste classification system in which one axis, related to ''requires permanent isolation,'' is associated with long-term risks from waste disposal and the other axis, related to ''highly radioactive,'' is associated with shorter-term risks due to high levels of decay heat and external radiation. We define wastes that require permanent isolation as wastes with concentrations of radionuclides exceeding the Class-C limits that are generally acceptable for near-surface land disposal, as specified in the US Nuclear Regulatory Commission's rulemaking 10 CFR Part 61 and its supporting documentation. HLW then is waste requiring permanent isolation that also is highly radioactive, and we define ''highly radioactive'' as a decay heat (power density) in the waste greater than 50 W/m/sup 3/ or an external radiation dose rate at a distance of 1 m from the waste greater than 100 rem/h (1 Sv/h), whichever is the more restrictive. This proposal also results in a definition of Transuranic (TRU) Waste and Equivalent as waste that requires permanent isolation but is not highly radioactive and a definition of low-level waste (LLW) as waste that does not require permanent isolation without regard to whether or not it is highly radioactive.

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

    SciTech Connect

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

    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.

  16. 10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation....

  17. 10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation....

  18. 10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation....

  19. 10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation....

  20. 10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation....

  1. Parametric Analyses of Heat Removal from High Level Waste Tanks

    SciTech Connect

    TRUITT, J.B.

    2000-06-05

    The general thermal hydraulics program GOTH-SNF was used to predict the thermal response of the waste in tanks 241-AY-102 and 241-AZ-102 when mixed by two 300 horsepower mixer pumps. This mixing was defined in terms of a specific waste retrieval scenario. Both dome and annulus ventilation system flow are necessary to maintain the waste within temperature control limits during the mixing operation and later during the sludge-settling portion of the scenario are defined.

  2. Criticality safety of high-level tank waste

    SciTech Connect

    Rogers, C.A.

    1995-07-01

    Radioactive waste containing low concentrations of fissile isotopes is stored in underground storage tanks on the Hanford Site in Washington State. The goal of criticality safety is to ensure that this waste remains subcritical into the indefinite future without supervision. A large ratio of solids to plutonium provides an effective way of ensuring a low plutonium concentration. Since the first waste discharge, a program of audits and appraisals has ensured that operations are conducted according to limits and controls applied to them. In addition, a program of surveillance and characterization maintains watch over waste after discharge.

  3. High-Level Radioactive Waste: Safe Storage and Ultimate Disposal.

    ERIC Educational Resources Information Center

    Dukert, Joseph M.

    Described are problems and techniques for safe disposal of radioactive waste. Degrees of radioactivity, temporary storage, and long-term permanent storage are discussed. Included are diagrams of estimated waste volumes to the year 2000 and of an artist's conception of a permanent underground disposal facility. (SL)

  4. STABILITY OF HIGH-LEVEL RADIOACTIVE WASTE FORMS

    EPA Science Inventory

    The objective of the proposed effort is to expand the development of solution models of complex waste glass systems that are predictive with regard to composition, phase separation, chemical activity, and volatility. The effort will also yield thermodynamic values for waste compo...

  5. FINAL REPORT. STABILITY OF HIGH-LEVEL WASTE FORMS

    EPA Science Inventory

    The objective of the proposed effort was to expand the development of solution models of complex waste glass systems that are predictive with regard to composition, phase separation, chemical activity, and volatility. The effort was to yield thermodynamic values for waste compone...

  6. Foreign programs for the storage of spent nuclear power plant fuels, high-level waste canisters and transuranic wastes

    SciTech Connect

    Harmon, K.M.; Johnson, A.B. Jr.

    1984-04-01

    The various national programs for developing and applying technology for the interim storage of spent fuel, high-level radioactive waste, and TRU wastes are summarized. Primary emphasis of the report is on dry storage techniques for uranium dioxide fuels, but data are also provided concerning pool storage.

  7. PLUTONIUM SOLUBILITY IN HIGH-LEVEL WASTE ALKALI BOROSILICATE GLASS

    SciTech Connect

    Marra, J.; Crawford, C.; Fox, K.; Bibler, N.

    2011-01-04

    The solubility of plutonium in a Sludge Batch 6 (SB6) reference glass and the effect of incorporation of Pu in the glass on specific glass properties were evaluated. A Pu loading of 1 wt % in glass was studied. Prior to actual plutonium glass testing, surrogate testing (using Hf as a surrogate for Pu) was conducted to evaluate the homogeneity of significant quantities of Hf (Pu) in the glass, determine the most appropriate methods to evaluate homogeneity for Pu glass testing, and to evaluate the impact of Hf loading in the glass on select glass properties. Surrogate testing was conducted using Hf to represent between 0 and 1 wt % Pu in glass on an equivalent molar basis. A Pu loading of 1 wt % in glass translated to {approx}18 kg Pu per Defense Waste Processing Facility (DWPF) canister, or about 10X the current allowed limit per the Waste Acceptance Product Specifications (2500 g/m{sup 3} of glass or about 1700 g/canister) and about 30X the current allowable concentration based on the fissile material concentration limit referenced in the Yucca Mountain Project License Application (897 g/m{sup 3}3 of glass or about 600 g Pu/canister). Based on historical process throughput data, this level was considered to represent a reasonable upper bound for Pu loading based on the ability to provide Pu containing feed to the DWPF. The task elements included evaluating the distribution of Pu in the glass (e.g. homogeneity), evaluating crystallization within the glass, evaluating select glass properties (with surrogates), and evaluating durability using the Product Consistency Test -- Method A (PCT-A). The behavior of Pu in the melter was evaluated using paper studies and corresponding analyses of DWPF melter pour samples.The results of the testing indicated that at 1 wt % Pu in the glass, the Pu was homogeneously distributed and did not result in any formation of plutonium-containing crystalline phases as long as the glass was prepared under 'well-mixed' conditions. The

  8. Hanford high-level waste evaporator/crystallizer corrosion evaluation

    SciTech Connect

    Ohl, P.C.; Carlos, W.C.

    1993-10-01

    The US Department of Energy, Hanford Site nuclear reservation, located in Southeastern Washington State, is currently home to 61 Mgal of radioactive waste stored in 177 large underground storage tanks. As an intermediate waste volume reduction, the 242-A Evaporator/Crystallizer processes waste solutions from most of the operating laboratories and plants on the Hanford Site. The waste solutions are concentrated in the Evaporator/Crystallizer to a slurry of liquid and crystallized salts. This concentrated slurry is returned to Hanford Site waste tanks at a significantly reduced volume. The Washington State Department of Ecology Dangerous Waste Regulations, WAC 173-393 require that a tank system integrity assessment be completed and maintained on file at the facility for all dangerous waste tank systems. This corrosion evaluation was performed in support of the 242-A Evaporator/Crystallizer Tank System Integrity Assessment Report. This corrosion evaluation provided a comprehensive compatibility study of the component materials and corrosive environments. Materials used for the Evaporator components and piping include austenitic stainless steels (SS) (primarily ASTM A240, Type 304L) and low alloy carbon steels (CS) (primarily ASTM A53 and A106) with polymeric or asbestos gaskets at flanged connections. Building structure and secondary containment is made from ACI 301-72 Structural Concrete for Buildings and coated with a chemically resistant acrylic coating system.

  9. SETTLING OF SPINEL IN A HIGH-LEVEL WASTE GLASS MELTER

    SciTech Connect

    Pavel Hrma; Pert Schill; Lubomir Nemec

    2002-01-07

    High-level nuclear waste is being vitrified, i.e., converted to a durable glass that can be stored in a safe repository for hundreds of thousands of years. Waste vitrification is accomplished in reactors called melters to which the waste is charged together with glass-forming additives. The mixture is electrically heated to a temperature as high as 1150 decrees C to create a melt that becomes glass on cooling.

  10. Corrosion issues in high-level nuclear waste containers

    NASA Astrophysics Data System (ADS)

    Asl, Samin Sharifi

    In this dissertation different aspects of corrosion and electrochemistry of copper, candidate canister material in Scandinavian high-level nuclear waste disposal program, including the thermodynamics and kinetics of the reactions that are predicted to occur in the practical system have been studied. A comprehensive thermodynamic study of copper in contact with granitic groundwater of the type and composition that is expected in the Forsmark repository in Sweden has been performed. Our primary objective was to ascertain whether copper would exist in the thermodynamically immune state in the repository, in which case corrosion could not occur and the issue of corrosion in the assessment of the storage technology would be moot. In spite of the fact that metallic copper has been found to exist for geological times in granitic geological formations, copper is well-known to be activated from the immune state to corrode by specific species that may exist in the environment. The principal activator of copper is known to be sulfur in its various forms, including sulfide (H2S, HS-, S2-), polysulfide (H2Sx, HSx -, Sx 2-), poly sulfur thiosulfate ( SxO3 2-), and polythionates (SxO6 2-). A comprehensive study of this aspect of copper chemistry has never been reported, and yet an understanding of this issue is vital for assessing whether copper is a suitable material for fabricating canisters for the disposal of HLNW. Our study identifies and explores those species that activate copper; these species include sulfur-containing entities as well as other, non-sulfur species that may be present in the repository. The effects of temperature, solution pH, and hydrogen pressure on the kinetics of the hydrogen electrode reaction (HER) on copper in borate buffer solution have been studied by means of steady-state polarization measurements, including electrochemical impedance spectroscopy (EIS). In order to obtain electrokinetic parameters, such as the exchange current density and the

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

    SciTech Connect

    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.

  12. High-level waste disposal, ethics and thermodynamics

    NASA Astrophysics Data System (ADS)

    Schwartz, Michael O.

    2008-06-01

    Moral philosophy applied to nuclear waste disposal can be linked to paradigmatic science. Simple thermodynamic principles tell us something about rightness or wrongness of our action. Ethical judgement can be orientated towards the chemical compatibility between waste container and geological repository. A container-repository system as close as possible to thermodynamic equilibrium is ethically acceptable. It aims at unlimited stability, similar to the stability of natural metal deposits within the Earth’s crust. The practicability of the guideline can be demonstrated.

  13. Boehmite Dissolution Studies Supporting High Level Waste Pretreatment - 9383

    SciTech Connect

    Peterson, Reid A.; Russell, Renee L.; Snow, Lanee A.

    2009-03-01

    Boehmite is present in significant quantities in several of the Hanford waste tanks. It has been proposed that the boehmite will be dissolved through caustic leaching in the Hanford Waste Treatment Plant currently under construction. Therefore, it is important to fully understand the nature of this dissolution so that the process can be deployed. This research determined the impact of primary control parameters on the boehmite dissolution rate. The impact of aluminate ion on the dissolution kinetics was determined. In addition, other parameters that impact boehmite dissolution, such as free hydroxide concentration and reaction temperature, were also assessed and used to develop a semi-empirical model of the boehmite dissolution process. The understanding derived from this work will be used as the basis to evaluate and improve the planned performance of the Hanford Waste Treatment plant. This work is the first in a series of programs aimed at demonstrating the Waste Treatment Plant dissolution process. This work will be used to develop a simulant of the boehmite-containing Hanford waste. That simulant will then be used in laboratory- and pilot-scale testing to demonstrate the Waste Treatment Plant pretreatment process in an integrated fashion.

  14. Microwave energy for post-calcination treatment of high-level nuclear wastes

    SciTech Connect

    Gombert, D.; Priebe, S.J.; Berreth, J.R.

    1980-01-01

    High-level radioactive wastes generated from nuclear fuel reprocessing require treatment for effective long-term storage. Heating by microwave energy is explored in processing of two possible waste forms: (1) drying of a pelleted form of calcined waste; and (2) vitrification of calcined waste. It is shown that residence times for these processes can be greatly reduced when using microwave energy rather than conventional heating sources, without affecting product properties. Compounds in the waste and in the glass frit additives couple very well with the 2.45 GHz microwave field so that no special microwave absorbers are necessary.

  15. High-performance gamma spectroscopy for equipment retrieval from Hanford high-level nuclear waste tanks

    NASA Astrophysics Data System (ADS)

    Troyer, Gary L.; Hillesand, K. E.; Goodwin, S. G.; Kessler, S. F.; Killian, E. W.; Legare, D.; Nelson, Joseph V., Jr.; Richard, R. F.; Nordquist, E. M.

    1999-01-01

    The cleanup of high level defense nuclear waste at the Hanford site presents several progressive challenges. Among these is the removal and disposal of various components from buried active waste tanks to allow new equipment insertion or hazards mitigation. A unique automated retrieval system at the tank provides for retrieval, high pressure washing, inventory measurement, and containment for disposal. Key to the inventory measurement is a three detector HPGe high performance gamma spectroscopy system capable of recovering data at up to ninety per cent saturation (200,000 counts per second). Data recovery is based on a unique embedded electronic pulser and specialized software to report the inventory. Each of the detectors have different shielding specified through Monte Carlo simulation with the MCNP program. This shielding provides performance over a dynamic range of eight orders of magnitude. System description, calibration issues and operational experiences are discussed.

  16. High-level waste processing at the Savannah River Site: An update

    SciTech Connect

    Marra, J.E.; Bennett, W.M.; Elder, H.H.; Lee, E.D.; Marra, S.L.; Rutland, P.L.

    1997-09-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) in Aiken, SC mg began immobilizing high-level radioactive waste in borosilicate glass in 1996. Currently, the radioactive glass is being produced as a ``sludge-only`` composition by combining washed high-level waste sludge with glass frit. The glass is poured in stainless steel canisters which will eventually be disposed of in a permanent, geological repository. To date, DWPF has produced about 100 canisters of vitrified waste. Future processing operations will, be based on a ``coupled`` feed of washed high-level waste sludge, precipitated cesium, and glass frit. This paper provides an update of the processing activities completed to date, operational/flowsheet problems encountered, and programs underway to increase production rates.

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

    SciTech Connect

    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.

  18. Hanford Site River Protection Project High-Level Waste Safe Storage and Retrieval

    SciTech Connect

    Aromi, E. S.; Raymond, R. E.; Allen, D. I.; Payne, M. A.; DeFigh-Price, C.; Kristofzski, J. G.; Wiegman, S. A.

    2002-02-25

    This paper provides an update from last year and describes project successes and issues associated with the management and work required to safely store, enhance readiness for waste feed delivery, and prepare for treated waste receipts for the approximately 53 million gallons of mixed and high-level waste currently in aging tanks at the Hanford Site. The Hanford Site is a 560 square-mile area in southeastern Washington State near Richland, Washington.

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

    SciTech Connect

    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.

  20. Composite quarterly technical report long-term high-level-waste technology, October-December 1981

    SciTech Connect

    Cornman, W.R.

    1982-06-01

    This document summarizes work performed at participating sites on the immobilization of high-level wastes from the chemical reprocessing of reactor fuels. The plan is to develop waste form alternatives for each of the three DOE sites (SRP, ICPP, and Hanford). Progress is reported in the following areas: waste preparation; fixation in glass, concrete, tailored ceramics, and coated particles; process and equipment development; and final handling. 12 figures, 19 tables. (DLC)

  1. Separating and Stabilizing Phosphate from High-Level Radioactive Waste: Process Development and Spectroscopic Monitoring

    SciTech Connect

    Lumetta, Gregg J.; Braley, Jenifer C.; Peterson, James M.; Bryan, Samuel A.; Levitskaia, Tatiana G.

    2012-05-09

    Removing phosphate from alkaline high-level waste sludges at the Department of Energy's Hanford Site in Washington State is necessary to increase the waste loading in the borosilicate glass waste form that will be used to immobilize the highly radioactive fraction of these wastes. We are developing a process which first leaches phosphate from the high-level waste solids with aqueous sodium hydroxide, and then isolates the phosphate by precipitation with calcium oxide. Tests with actual tank waste confirmed that this process is an effective method of phosphate removal from the sludge and offers an additional option for managing the phosphorus in the Hanford tank waste solids. The presence of vibrationally active species, such as nitrate and phosphate ions, in the tank waste processing streams makes the phosphate removal process an ideal candidate for monitoring by Raman or infrared spectroscopic means. As a proof-of-principle demonstration, Raman and Fourier transform infrared (FTIR) spectra were acquired for all phases during a test of the process with actual tank waste. Quantitative determination of phosphate, nitrate, and sulfate in the liquid phases was achieved by Raman spectroscopy, demonstrating the applicability of Raman spectroscopy for the monitoring of these species in the tank waste process streams.

  2. Separating and stabilizing phosphate from high-level radioactive waste: process development and spectroscopic monitoring.

    PubMed

    Lumetta, Gregg J; Braley, Jenifer C; Peterson, James M; Bryan, Samuel A; Levitskaia, Tatiana G

    2012-06-01

    Removing phosphate from alkaline high-level waste sludges at the Department of Energy's Hanford Site in Washington State is necessary to increase the waste loading in the borosilicate glass waste form that will be used to immobilize the highly radioactive fraction of these wastes. We are developing a process which first leaches phosphate from the high-level waste solids with aqueous sodium hydroxide, and then isolates the phosphate by precipitation with calcium oxide. Tests with actual tank waste confirmed that this process is an effective method of phosphate removal from the sludge and offers an additional option for managing the phosphorus in the Hanford tank waste solids. The presence of vibrationally active species, such as nitrate and phosphate ions, in the tank waste processing streams makes the phosphate removal process an ideal candidate for monitoring by Raman or infrared spectroscopic means. As a proof-of-principle demonstration, Raman and Fourier transform infrared (FTIR) spectra were acquired for all phases during a test of the process with actual tank waste. Quantitative determination of phosphate, nitrate, and sulfate in the liquid phases was achieved by Raman spectroscopy, demonstrating the applicability of Raman spectroscopy for the monitoring of these species in the tank waste process streams. PMID:22571620

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

    SciTech Connect

    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. Estimation of Failure Frequency for Type I and II High Level Waste Tanks

    SciTech Connect

    Subramanian, K.H.

    2001-05-15

    The failure frequency of Type I and Type II High Level Waste tanks was calculated. The degradation mechanism that could lead to large break failure and the credits taken for steps taken to prevent large break failure were considered.

  5. 75 FR 61228 - Board Meeting: Technical Lessons Gained From High-Level Nuclear Waste Disposal Efforts

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-04

    ... REVIEW BOARD Board Meeting: Technical Lessons Gained From High-Level Nuclear Waste Disposal Efforts... 1987, the U.S. Nuclear Waste Technical Review Board will meet in Dulles, Virginia, on October 26, 2010... technical issues and to review the technical validity of DOE activities related to implementing the...

  6. Chem I Supplement. Chemistry Related to Isolation of High-Level Nuclear Waste.

    ERIC Educational Resources Information Center

    Hoffman, Darleane C.; Choppin, Gregory R.

    1986-01-01

    Discusses some of the problems associated with the safe disposal of high-level nuclear wastes. Describes several waste disposal plans developed by various nations. Outlines the multiple-barrier concept of isolation in deep geological questions associated with the implementation of such a method. (TW)

  7. STUDIES RELATED TO CHEMICAL MECHANISMS OF GAS FORMATION IN HANFORD HIGH-LEVEL WASTES

    EPA Science Inventory

    The objective of this work is to develop a more detailed mechanistic understanding of the thermal reactions that lead to gas production in certain high-level waste storage tanks at the Hanford, Washington site. Prediction of the combustion hazard for these wastes and engineering ...

  8. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  9. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  10. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  11. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

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

    SciTech Connect

    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 canisters. The intended use of this analysis is to

  13. In-tank pretreatment of high-level tank wastes: The SIPS system

    SciTech Connect

    Reich, M.; Powell, J.; Barletta, R.

    1996-03-01

    A new approach, termed SIPS (Small In-Tank Processing System), that enables the in-tank processing and separation of high-level tank wastes into high-level waste (HLW) and low-level waste (LLW) streams that are suitable for vitrification, is described. Presently proposed pretreatment systems, such as enhanced sludge washing (ESW) and TRUEX, require that the high-level tank wastes be retrieved and pumped to a large, centralized processing facility, where the various waste components are separated into a relatively small, radioactively concentrated stream (HLW), and a relatively large, predominantly non-radioactive stream (LLW). In SIPS, a small process module, typically on the order of 1 meter in diameter and 4 meters in length, is inserted into a tank. During a period of approximately six months, it processes the solid/liquid materials in the tank, separating them into liquid HLW and liquid LLW output streams that are pumped away in two small diameter (typically 3 cm o.d.) pipes. The SIPS concept appears attractive for pretreating high level wastes, since it would: (1) process waste in-situ in the tanks, (2) be cheaper and more reliable than a larger centralized facility, (3) be quickly demonstrable at full scale, (4) have less technical risk, (5) avoid having to transfer unstable slurries for long distances, and (6) be simple to decommission and dispose of. Further investigation of the SIPS concept appears desirable, including experimental testing and development of subscale demonstration units.

  14. Long-Term High-Level Defense-Waste technology

    NASA Astrophysics Data System (ADS)

    1982-07-01

    In the residual liquid solidification effort, the primary alternative studied is the wiped film evaporator approach to solidifying salt well pumped liquids and returning the molten material to single shell tanks for microwave final stabilization to a hard dry product. Both systems analysis and experimental work are proceeding to evaluate this approach. The primary alternative for in situ stabilization of in-tank wastes is microwave drying of wet salt cake and unpumped sludges. Experimental work was successfully conducted on a 1/12 scale tank containing wet synthetic salt cake. Related systems analysis of a full scale system was initiated.

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

    SciTech Connect

    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.

  16. The radiation characteristics of the transport packages with vitrified high-level waste

    NASA Astrophysics Data System (ADS)

    Bogatov, S. A.; Mitenkova, E. F.; Novikov, N. V.

    2015-12-01

    The calculation method of neutron yield in the (α, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron yield in the (α, n) reaction. For vitrified high-level waste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified highlevel waste are estimated.

  17. The radiation characteristics of the transport packages with vitrified high-level waste

    SciTech Connect

    Bogatov, S. A.; Mitenkova, E. F. Novikov, N. V.

    2015-12-15

    The calculation method of neutron yield in the (α, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron yield in the (α, n) reaction. For vitrified high-level waste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified highlevel waste are estimated.

  18. Replacement inhibitors for high level waste tank cooling coils

    SciTech Connect

    Wiersma, B.J.; Hsu, T.C.

    1996-10-01

    Sodium chromate has been an effective corrosion inhibitor for the cooling coil systems in Savannah River Site (SRS) waste tanks for over 40 years. Due to age and operating history, cooling coils occasionally fail allowing chromated water to leak into the environment. The costs of reporting and cleaning up chromate spills became significant enough for SRS to consider alternate inhibitors. Confirmatory tests were performed to assess the effectiveness of three alternative corrosion inhibitor systems for the waste tank cooling water systems: (1) sodium molybdate (250 ppm as Mo)/sodium hydroxide (pH 10), (2) sodium molybdate (50 ppm as Mo)/sodium silicate (50 ppm as Si), and (3) sodium nitrite (500 ppm)/ sodium hydroxide (0.01 M). The tests were conducted under stagnant conditions to simulate a worst-case scenario. The results showed that these inhibitors were as effective as chromate at minimizing general corrosion at solution temperatures between 30--70 C. However, the initiation of localized attack in crevice regions, in solutions containing the alternative inhibitors at 70 C, was observed. Also, for the nitrite and the molybdate systems to be effective, suitable biocide is needed. On the other hand, interval coupon tests showed that the molybdate inhibitor systems prevented significant propagation of the localized attack.

  19. Yucca Mountain, Nevada - A proposed geologic repository for high-level radioactive waste

    USGS Publications Warehouse

    Levich, R.A.; Stuckless, J.S.

    2006-01-01

    Yucca Mountain in Nevada represents the proposed solution to what has been a lengthy national effort to dispose of high-level radioactive waste, waste which must be isolated from the biosphere for tens of thousands of years. This chapter reviews the background of that national effort and includes some discussion of international work in order to provide a more complete framework for the problem of waste disposal. Other chapters provide the regional geologic setting, the geology of the Yucca Mountain site, the tectonics, and climate (past, present, and future). These last two chapters are integral to prediction of long-term waste isolation. ?? 2007 Geological Society of America. All rights reserved.

  20. Yucca Mountain, Nevada - A Proposed Geologic Repository for High-Level Radioactive Waste (Volume 1) Introduction

    SciTech Connect

    R.A. Levich; J.S. Stuckless

    2006-09-25

    Yucca Mountain in Nevada represents the proposed solution to what has been a lengthy national effort to dispose of high-level radioactive waste, waste which must be isolated from the biosphere for tens of thousands of years. This chapter reviews the background of that national effort and includes some discussion of international work in order to provide a more complete framework for the problem of waste disposal. Other chapters provide the regional geologic setting, the geology of the Yucca Mountain site, the tectonics, and climate (past, present, and future). These last two chapters are integral to prediction of long-term waste isolation.

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

    SciTech Connect

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

    1999-05-19

    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.

  2. Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

    DOEpatents

    Boatner, Lynn A.; Sales, Brian C.

    1989-01-01

    Lead-iron phosphate glasses containing a high level of Fe.sub.2 O.sub.3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90.degree. C., with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10.sup.2 to 10.sup.3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe.sub.2 O.sub.3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800.degree. C., since they exhibit very low melt viscosities in the 800.degree. to 1050.degree. C. temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550.degree. C. and are not adversely affected by large doses of gamma radiation in H.sub.2 O at 135.degree. C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms.

  3. Overview of European concepts for high-level waste and spent fuel disposal with special reference waste container corrosion

    NASA Astrophysics Data System (ADS)

    Bennett, D. G.; Gens, R.

    2008-09-01

    This paper provides a brief overview of current repository and engineered barrier system (EBS) designs in selected high-level waste (HLW) and spent fuel (SF) disposal concepts from European countries, with special reference to key metallic waste containers and corrosion processes. The paper discusses assessments of copper, iron and steel container corrosion behaviour under the environmental conditions expected, given likely repository host rocks and groundwaters, and comments on the significance of corrosion processes, the choice of waste container materials, and areas of research. Most of the HLW and/or SF disposal programmes in European countries are pursuing disposal options in which the primary waste container is designed, in conjunction with the surrounding EBS materials, to provide complete containment of the waste for at least the period when temperatures in the disposal system are significantly raised by radioactive decay.

  4. On-site storage of high level nuclear waste: attitudes and perceptions of local residents.

    PubMed

    Bassett, G W; Jenkins-Smith, H C; Silva, C

    1996-06-01

    No public policy issue has been as difficult as high-level nuclear waste. Debates continue regarding Yucca Mountain as a disposal site, and-more generally-the appropriateness of geologic disposal and the need to act quickly. Previous research has focused on possible social, political, and economic consequences of a facility in Nevada. Impacts have been predicted to be potentially large and to emanate mainly from stigmatization of the region due to increased perceptions of risk. Analogous impacts from leaving waste at power plants have been either ignored or assumed to be negligible. This paper presents survey results on attitudes of residents in three counties where nuclear waste is currently stored. Topics include perceived risk, knowledge of nuclear waste and radiation, and impacts on jobs, tourism, and housing values from leaving waste on site. Results are similar to what has been reported for Nevada; the public is concerned about possible adverse effects from on-site storage of waste. PMID:8693158

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

    SciTech Connect

    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.

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

    SciTech Connect

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

    1994-05-01

    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. PSA results for Hanford high level waste Tank 101-SY

    SciTech Connect

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

    1993-10-01

    Los Alamos National Laboratory has performed a comprehensive probabilistic safety assessment (PSA) that includes consideration of external events for 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 (``burps``) a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed 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 underway 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. This PSA for Tank 101-SY, which did not consider the pump experiment or future tank-remediation activities, involved three distinct tasks. First, the accident sequence analysis identified and quantified those potential accidents whose consequences result in tank material release. Second, characteristics and release paths for the airborne and liquid radioactive source terms were determined. Finally, the consequences, primarily onsite and offsite potential health effects resulting from radionuclide release, were estimated, and overall risk curves were constructed. An overview of each of these tasks and a summary of the overall results of the analysis are presented in the following sections.

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

    SciTech Connect

    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.

  9. The HRA/Solarium Project: Processing of Widely Varying High- and Medium-Level Waste

    SciTech Connect

    Willems, M.; Luycx, P.; Gilis, R.; Belgoprocess; Renard, Cl.; Reyniers, H.; Cuchet, J. M.

    2003-02-26

    Starting in 2003, Belgoprocess will proceed with the treatment and conditioning of some 200 m{sup 3} of widely varying high- and medium-level waste from earlier research and development work, to meet standard acceptance criteria for later disposal. The gross volume of primary and secondary packages amounts to 2,600 m{sup 3}. The waste has been kept in decay storage for up to 30 years. The project was started in 1997. Operation of the various processing facilities will take 7-8 years. The overall volume of conditioned waste will be of the order of 800 m{sup 3}. All conditioned waste will be stored in appropriate storage facilities onsite. At present (November, 2002), a new processing facility has been constructed, the functional tests of the equipment have been performed and the startup phase has been started. Several cells of the Pamela vitrification facility onsite will be adapted for the treatment of high-level and highly a-contaminated waste; low-level a/a waste will be treated in the existing facility for super compaction and conditioning by embedding into cement (CILVA). The bulk of these waste, of which 95% are solids, the remainder consisting of mainly solidified liquids, have been produced between 1967 and 1988. They originate from various research programs and reactor operation at the Belgian nuclear energy research centre SCK CEN, isotope production, decontamination and dismantling operations.

  10. Chemical Environment at Waste Package Surfaces in a High-Level Radioactive Waste Repository

    SciTech Connect

    Carroll, S; Alai, M; Craig, L; Gdowski, G; Hailey, P; Nguyen, Q A; Rard, J; Staggs, K; Sutton, M; Wolery, T

    2005-05-26

    We have conducted a series of deliquescence, boiling point, chemical transformation, and evaporation experiments to determine the composition of waters likely to contact waste package surfaces over the thermal history of the repository as it heats up and cools back down to ambient conditions. In the above-boiling period, brines will be characterized by high nitrate to chloride ratios that are stable to higher temperatures than previously predicted. This is clearly shown for the NaCl-KNO{sub 3} salt system in the deliquescence and boiling point experiments in this report. Our results show that additional thermodynamic data are needed in nitrate systems to accurately predict brine stability and composition due to salt deliquescence in dust deposited on waste package surfaces. Current YMP models capture dry-out conditions but not composition for NaCl-KNO{sub 3} brines, and they fail to predict dry-out conditions for NaCl-KNO{sub 3}-NaNO{sub 3} brines. Boiling point and deliquescence experiments are needed in NaCl-KNO{sub 3}-NaNO{sub 3} and NaCl-KNO{sub 3}-NaNO{sub 3}-Ca(NO{sub 3}){sub 2} systems to directly determine dry-out conditions and composition, because these salt mixtures are also predicted to control brine composition in the above-boiling period. Corrosion experiments are needed in high temperature and high NO{sub 3}:Cl brines to determine if nitrate inhibits corrosion in these concentrated brines at temperatures above 160 C. Chemical transformations appear to be important for pure calcium- and magnesium-chloride brines at temperatures greater than 120 C. This stems from a lack of acid gas volatility in NaCl/KNO{sub 3} based brines and by slow CO{sub 2}(g) diffusion in alkaline brines. This suggests that YMP corrosion models based on bulk solution experiments over the appropriate composition, temperature, and relative humidity range can be used to predict corrosion in thin brine films formed by salt deliquescence. In contrast to the above-boiling period, the

  11. Long-Term Waste Package Degradation Studies at the Yucca Mountain Potential High-Level Nuclear Waste Repository

    SciTech Connect

    Mon, K. G.; Bullard, B. E.; Longsine, D. E.; Mehta, S.; Lee, J. H.; Monib, A. M.

    2002-02-26

    The Site Recommendation (SR) process for the potential repository for spent nuclear fuel (SNF) and high-level nuclear waste (HLW) at Yucca Mountain, Nevada is underway. Fulfillment of the requirements for substantially complete containment of the radioactive waste emplaced in the potential repository and subsequent slow release of radionuclides from the Engineered Barrier System (EBS) into the geosphere will rely on a robust waste container design, among other EBS components. Part of the SR process involves sensitivity studies aimed at elucidating which model parameters contribute most to the drip shield and waste package degradation characteristics. The model parameters identified included (a) general corrosion rate model parameters (temperature-dependence and uncertainty treatment), and (b) stress corrosion cracking (SCC) model parameters (uncertainty treatment of stress and stress intensity factor profiles in the Alloy 22 waste package outer barrier closure weld regions, the SCC initiation stress threshold, and the fraction of manufacturing flaws oriented favorably for through-wall penetration by SCC). These model parameters were reevaluated and new distributions were generated. Also, early waste package failures due to improper heat treatment were added to the waste package degradation model. The results of these investigations indicate that the waste package failure profiles are governed by the manufacturing flaw orientation model parameters and models used.

  12. GLASS FORMULATION DEVELOPMENT TO SUPPORT MELTER TESTING TO DEMONSTRATE ENHANCED HIGH LEVEL WASTE THROUGHPUT

    SciTech Connect

    Marra, J; David Peeler, D; Tommy Edwards, T; Kevin Fox, K; Amanda Youchak, A; James Gillam, J

    2007-08-17

    The U.S. Department of Energy (DOE) is currently processing high-level waste (HLW) through a Joule-heated melter (JHM) at the Savannah River Site (SRS) and plans to vitrify HLW and Low activity waste (LAW) at the Hanford Site. Over the past few years at the DWPF, work has concentrated on increasing waste throughput. These efforts are continuing with an emphasis on high alumina content feeds. High alumina feeds have presented specific challenges for the JHM technology regarding the ability to increase waste loading yet still maintain product quality and adequate throughput. Alternatively, vitrification technology innovations are also being investigated as a means to increase waste throughput. The Cold Crucible Induction Melter (CCIM) technology affords the opportunity for higher vitrification process temperatures as compared to the current reference JHM technology. Higher process temperatures may allow for higher waste loading and higher melt rate. Glass formulation testing to support melter demonstration testing was recently completed. This testing was specifically aimed at high alumina concentration wastes. Glass composition property models were utilized as a guide for formulation development. Both CCIM and JHM testing will be conducted so glass formulation testing was targeted at both technologies with a goal to significantly increase waste loading without compromising product quality.

  13. Special Analysis for Disposal of High-Concentration I-129 Waste in the Low-Activity Waste Vaults at the E-Area Low-Level Waste Facility

    SciTech Connect

    Collard, L.B.

    2003-02-14

    This Special Analysis (SA) addresses disposal of wastes containing a high concentration of I-129 in the Low Activity Waste (LAW) Vaults at the Savannah River Site E-Area Low-Level Waste Facility. These wastes were analyzed by computer modeling incorporating a laboratory-measured I-129 Kd. A Kd represents partitioning of a contaminant between solid particles (i.e., the waste) and liquid that can transport the contaminant. These wastes exhibited high Kds indicating a slow release rate that typically is manifested as low aquifer concentrations and high vault-inventory limits. This SA applies to ER and ETF wastes and any type of future waste that contains a high concentration of I-129. This SA was prepared to meet the requirements of the U.S. Department of Energy Order 435.1 (DOE 1999). Because the Performance Assessment (PA, McDowell-Boyer, et al., 2000) analyzes generic wastes, this report focuses on wastes with a high I-129 concentration. Common information from the PA is not duplicate d in this report.

  14. Potential for radiation damage to carbon steel storage tanks for high level radioactive waste

    SciTech Connect

    Caskey, G.R. Jr.; Sindelar, R.L.; Thomas, J.K.

    1993-07-30

    A low intensity radiation field is generated by the high level waste that is stored within carbon steel lined tanks at the Savannah River Site (SRS). The highest level of radiation damage to the tank walls from gamma and spontaneous neutron emissions is estimated to be less than 1.0E-6 displacements per atom (DPA) for a 100 year exposure to fresh, ``high heat`` SRS waste assuming continuous replenishment of the radionuclides. This damage level is below the limit for measurable radiation damage to the mechanical properties of carbon steel. Structural assessment of tanks for storage of high level waste may be based on nominal or code values of the mechanical properties of the steels from which the tanks were constructed.

  15. Separation of strontium-90 from Hanford high-level radioactive waste

    SciTech Connect

    Lumetta, G.J.; Wagner, M.J.; Jones, E.O.

    1993-10-01

    Current guidelines for disposing of high-level radioactive wastes stored in underground tanks at the US Department of Energy`s Hanford Site call for vitrifying high-level waste (HLW) in borosilicate glass and disposing the glass canisters in a deep geologic repository. Disposition of the low-level waste (LLW) is yet to be determined, but it will likely be immobilized in a glass matrix and disposed of on site. To lower the radiological risk associated with the LLW form, methods are being developed to separate {sup 90}Sr from the bulk waste material so this isotope can be routed to the HLW stream. A solvent extraction method is being investigated to separate {sup 90}Sr from acid-dissolved Hanford tank wastes. Results of experiments with actual tank waste indicate that this method can be used to achieve separation of {sup 90}Sr from the bulk waste components. Greater than 99% of the {sup 90}Sr was removed from an acidic dissolved sludge solution by extraction with di-tbutylcyclohexano-18-crown-6 in 1-octanol (the SREX process). The major sludge components were not extracted.

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

    SciTech Connect

    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.

  17. Minor component study for simulated high-level nuclear waste glasses (Draft)

    SciTech Connect

    Li, H.; Langowskim, M.H.; Hrma, P.R.; Schweiger, M.J.; Vienna, J.D.; Smith, D.E.

    1996-02-01

    Hanford Site single-shell tank (SSI) and double-shell tank (DSI) wastes are planned to be separated into low activity (or low-level waste, LLW) and high activity (or high-level waste, HLW) fractions, and to be vitrified for disposal. Formulation of HLW glass must comply with glass processibility and durability requirements, including constraints on melt viscosity, electrical conductivity, liquidus temperature, tendency for phase segregation on the molten glass surface, and chemical durability of the final waste form. A wide variety of HLW compositions are expected to be vitrified. In addition these wastes will likely vary in composition from current estimates. High concentrations of certain troublesome components, such as sulfate, phosphate, and chrome, raise concerns about their potential hinderance to the waste vitrification process. For example, phosphate segregation in the cold cap (the layer of feed on top of the glass melt) in a Joule-heated melter may inhibit the melting process (Bunnell, 1988). This has been reported during a pilot-scale ceramic melter run, PSCM-19, (Perez, 1985). Molten salt segregation of either sulfate or chromate is also hazardous to the waste vitrification process. Excessive (Cr, Fe, Mn, Ni) spinel crystal formation in molten glass can also be detrimental to melter operation.

  18. High Level Waste Remote Handling Equipment in the Melter Cave Support Handling System at the Hanford Waste Treatment Plant

    SciTech Connect

    Bardal, M.A.; Darwen, N.J.

    2008-07-01

    Cold war plutonium production led to extensive amounts of radioactive waste stored in tanks at the Department of Energy's (DOE) Hanford site. Bechtel National, Inc. is building the largest nuclear Waste Treatment Plant in the world located at the Department of Energy's Hanford site to immobilize the millions of gallons of radioactive waste. The site comprises five main facilities; Pretreatment, High Level Waste vitrification, Low Active Waste vitrification, an Analytical Lab and the Balance of Facilities. The pretreatment facilities will separate the high and low level waste. The high level waste will then proceed to the HLW facility for vitrification. Vitrification is a process of utilizing a melter to mix molten glass with radioactive waste to form a stable product for storage. The melter cave is designated as the High Level Waste Melter Cave Support Handling System (HSH). There are several key processes that occur in the HSH cell that are necessary for vitrification and include: feed preparation, mixing, pouring, cooling and all maintenance and repair of the process equipment. Due to the cell's high level radiation, remote handling equipment provided by PaR Systems, Inc. is required to install and remove all equipment in the HSH cell. The remote handling crane is composed of a bridge and trolley. The trolley supports a telescoping tube set that rigidly deploys a TR 4350 manipulator arm with seven degrees of freedom. A rotating, extending, and retracting slewing hoist is mounted to the bottom of the trolley and is centered about the telescoping tube set. Both the manipulator and slewer are unique to this cell. The slewer can reach into corners and the manipulator's cross pivoting wrist provides better operational dexterity and camera viewing angles at the end of the arm. Since the crane functions will be operated remotely, the entire cell and crane have been modeled with 3-D software. Model simulations have been used to confirm operational and maintenance

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

    SciTech Connect

    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

  20. Liquidus Temperature and Primary Crystallization Phases in High-Zirconia High-Level Waste Borosilicate Glasses

    SciTech Connect

    Plaisted, Trevor J.; Hrma, Pavel R.; Vienna, John D.; Jiricka, Antonin

    1999-12-09

    Liquidus temperature (TL) studies of high-Zr high-level waste (HLW) borosilicate glasses have identified three primary phases: baddelyite (ZrO2), zircon (ZrSiO4), and alkali-zirconium silicates, such as parakeldyshite (Na2ZrSi2O7). Using published TL data for HLW glasses with these primary phases, we have computed partial specific TLs for major glass components. On the Na2O-SiO2-ZrO2 submixture, we have determined approximate positions of the boundaries between the baddelyite, zircon, and parakeldyshite primary phase fields. The maximum that can dissolve at 1150?C in a borosilicate HLW glass subjected to common processability and acceptability constraints appears to be 16.5 mass% ZrO2.

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

    SciTech Connect

    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.

  2. Thermal analysis of Yucca Mountain commercial high-level waste packages

    SciTech Connect

    Altenhofen, M.K.; Eslinger, P.W.

    1992-10-01

    The thermal performance of commercial high-level waste packages was evaluated on a preliminary basis for the candidate Yucca Mountain repository site. The purpose of this study is to provide an estimate for waste package component temperatures as a function of isolation time in tuff. Several recommendations are made concerning the additional information and modeling needed to evaluate the thermal performance of the Yucca Mountain repository system.

  3. Optimization of waste loading in high-level glass in the presence of uncertainty

    SciTech Connect

    Hoza, M.; Fann, G.I.; Hopkins, D.F.

    1995-02-01

    Hanford high-level liquid waste will be converted into a glass form for long-term storage. The glass must meet certain constraints on its composition and properties in order to have desired properties for processing (e.g., electrical conductivity, viscosity, and liquidus temperature) and acceptable durability for long-term storage. The Optimal Waste Loading (OWL) models, based on rigorous mathematical optimization techniques, have been developed to minimize the number of glass logs required and determine glass-former compositions that will produce a glass meeting all relevant constraints. There is considerable uncertainty in many of the models and data relevant to the formulation of high-level glass. In this paper, we discuss how we handle uncertainty in the glass property models and in the high-level waste composition to the vitrification process. Glass property constraints used in optimization are inequalities that relate glass property models obtained by regression analysis of experimental data to numerical limits on property values. Therefore, these constraints are subject to uncertainty. The sampling distributions of the regression models are used to describe the uncertainties associated with the constraints. The optimization then accounts for these uncertainties by requiring the constraints to be satisfied within specified confidence limits. The uncertainty in waste composition is handled using stochastic optimization. Given means and standard deviations of component masses in the high-level waste stream, distributions of possible values for each component are generated. A series of optimization runs is performed; the distribution of each waste component is sampled for each run. The resultant distribution of solutions is then statistically summarized. The ability of OWL models to handle these forms of uncertainty make them very useful tools in designing and evaluating high-level waste glasses formulations.

  4. The Savannah River Site Replacement High Level Radioactive Waste Evaporator Project

    SciTech Connect

    Presgrove, S.B.

    1992-08-01

    The Replacement High Level Waste Evaporator Project was conceived in 1985 to reduce the volume of the high level radioactive waste Process of the high level waste has been accomplished up to this time using Bent Tube type evaporators and therefore, that type evaporator was selected for this project. The Title I Design of the project was 70% completed in late 1990. The Department of Energy at that time hired an independent consulting firm to perform a complete review of the project. The DOE placed a STOP ORDER on purchasing the evaporator in January 1991. Essentially, no construction was to be done on this project until all findings and concerns dealing with the type and design of the evaporator are resolved. This report addresses two aspects of the DOE design review; (1) Comparing the Bent Tube Evaporator with the Forced Circulation Evaporator, (2) The design portion of the DOE Project Review - concentrated on the mechanical design properties of the evaporator. 1 ref.

  5. The Savannah River Site Replacement High Level Radioactive Waste Evaporator Project

    SciTech Connect

    Presgrove, S.B. )

    1992-01-01

    The Replacement High Level Waste Evaporator Project was conceived in 1985 to reduce the volume of the high level radioactive waste Process of the high level waste has been accomplished up to this time using Bent Tube type evaporators and therefore, that type evaporator was selected for this project. The Title I Design of the project was 70% completed in late 1990. The Department of Energy at that time hired an independent consulting firm to perform a complete review of the project. The DOE placed a STOP ORDER on purchasing the evaporator in January 1991. Essentially, no construction was to be done on this project until all findings and concerns dealing with the type and design of the evaporator are resolved. This report addresses two aspects of the DOE design review; (1) Comparing the Bent Tube Evaporator with the Forced Circulation Evaporator, (2) The design portion of the DOE Project Review - concentrated on the mechanical design properties of the evaporator. 1 ref.

  6. Heat transfer analysis of the geologic disposal of spent fuel and high level waste storage canisters

    NASA Astrophysics Data System (ADS)

    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 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. 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.

  7. HIGH TEMPERATURE TREATMENT OF INTERMEDIATE-LEVEL RADIOACTIVE WASTES - SIA RADON EXPERIENCE

    SciTech Connect

    Sobolev, I.A.; Dmitriev, S.A.; Lifanov, F.A.; Kobelev, A.P.; Popkov, V.N.; Polkanov, M.A.; Savkin, A.E.; Varlakov, A.P.; Karlin, S.V.; Stefanovsky, S.V.; Karlina, O.K.; Semenov, K.N.

    2003-02-27

    This review describes high temperature methods of low- and intermediate-level radioactive waste (LILW) treatment currently used at SIA Radon. Solid and liquid organic and mixed organic and inorganic wastes are subjected to plasma heating in a shaft furnace with formation of stable leach resistant slag suitable for disposal in near-surface repositories. Liquid inorganic radioactive waste is vitrified in a cold crucible based plant with borosilicate glass productivity up to 75 kg/h. Radioactive silts from settlers are heat-treated at 500-700 0C in electric furnace forming cake following by cake crushing, charging into 200 L barrels and soaking with cement grout. Various thermochemical technologies for decontamination of metallic, asphalt, and concrete surfaces, treatment of organic wastes (spent ion-exchange resins, polymers, medical and biological wastes), batch vitrification of incinerator ashes, calcines, spent inorganic sorbents, contaminated soil, treatment of carbon containing 14C nuclide, reactor graphite, lubricants have been developed and implemented.

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

    SciTech Connect

    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.

  9. An analysis of the technical status of high level radioactive waste and spent fuel management systems

    NASA Technical Reports Server (NTRS)

    English, T.; Miller, C.; Bullard, E.; Campbell, R.; Chockie, A.; Divita, E.; Douthitt, C.; Edelson, E.; Lees, L.

    1977-01-01

    The technical status of the old U.S. mailine program for high level radioactive nuclear waste management, and the newly-developing program for disposal of unreprocessed spent fuel was assessed. The method of long term containment for both of these waste forms is considered to be deep geologic isolation in bedded salt. Each major component of both waste management systems is analyzed in terms of its scientific feasibility, technical achievability and engineering achievability. The resulting matrix leads to a systematic identification of major unresolved technical or scientific questions and/or gaps in these programs.

  10. Workshop on the role of natural analogs in geologic disposal of high-level nuclear waste

    SciTech Connect

    Murphy, W.M.; Kovach, L.A.

    1995-09-01

    A workshop on the Role of Natural Analogs in Geologic Disposal of High-Level Nuclear Waste (HLW) was held in San Antonio, Texas, on July 22-25, 1991. It was sponsored by the US Nuclear Regulatory Commission (NRC) and the Center for Nuclear Waste Regulatory Analyses (CNWRA). Invitations to the workshop were extended to a large number of individuals with a variety of technical and professional interests related to geologic disposal of nuclear waste and natural analog studies. The objective of the workshop was to examine the role of natural analog studies in performance assessment, site characterization, and prioritization of research related to geologic disposal of HLW.

  11. MICROSTRUCTURAL PROPERTIES OF HIGH-LEVEL WASTE CONCENTRATES AND GELS WITH RAMAN AND INFRARED SPECTROSCOPIES

    EPA Science Inventory

    Nearly half of the high level radioactive waste stored at Hanford is composed of highly alkaline concentrates referred to as either salt cakes or Double-Shell Slurry (DSS), depending on their compositions and processing histories. The major components of these concentrates are wa...

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

    SciTech Connect

    Christian, J. H.

    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. Polysiloxane Encapsulation of High Level Calcine Waste for Transportation or Disposal

    SciTech Connect

    Loomis, Guy George

    2000-03-01

    This report presents the results of an experimental study investigating the potential uses for silicon-polymer encapsulation of High Level Calcine Waste currently stored within the Idaho Nuclear Technology and Engineering Center (INTEC) at the Idaho National Engineering and Environmental Laboratory (INEEL). The study investigated two different applications of silicon polymer encapsulation. One application uses silicon polymer to produce a waste form suitable for disposal at a High Level Radioactive Waste Disposal Facility directly, and the other application encapsulates the calcine material for transportation to an offsite melter for further processing. A simulated waste material from INTEC, called pilot scale calcine, which contained hazardous materials but no radioactive isotopes was used for the study, which was performed at the University of Akron under special arrangement with Orbit Technologies, the originators of the silicon polymer process called Polymer Encapsulation Technology (PET). This document first discusses the PET process, followed by a presentation of past studies involving PET applications to waste problems. Next, the results of an experimental study are presented on encapsulation of the INTEC calcine waste as it applies to transportation or disposal of calcine waste. Results relating to long-term disposal include: 1) a characterization of the pilot calcine waste; 2) Toxicity Characteristic Leaching Procedure (TCLP) testing of an optimum mixture of pilot calcine, polysiloxane and special additives; and, 3) Material Characterization Center testing MCC-1P evaluation of the optimum waste form. Results relating to transportation of the calcine material for a mixture of maximum waste loading include: compressive strength testing, 10-m drop test, melt testing, and a Department of Transportation (DOT) oxidizer test.

  14. Silicon-Polymer Encapsulation of High-Level Calcine Waste for Transportation or Disposal

    SciTech Connect

    G. G. Loomis; C. M. Miller; J. A. Giansiracusa; R. Kimmel; S. V. Prewett

    2000-01-01

    This report presents the results of an experimental study investigating the potential uses for silicon-polymer encapsulation of High Level Calcine Waste currently stored within the Idaho Nuclear Technology and Engineering Center (INTEC) at the Idaho National Engineering and Environmental Laboratory (INEEL). The study investigated two different applications of silicon polymer encapsulation. One application uses silicon polymer to produce a waste form suitable for disposal at a High Level Radioactive Waste Disposal Facility directly, and the other application encapsulates the calcine material for transportation to an offsite melter for further processing. A simulated waste material from INTEC, called pilot scale calcine, which contained hazardous materials but no radioactive isotopes was used for the study, which was performed at the University of Akron under special arrangement with Orbit Technologies, the originators of the silicon polymer process called Polymer Encapsulation Technology (PET). This document first discusses the PET process, followed by a presentation of past studies involving PET applications to waste problems. Next, the results of an experimental study are presented on encapsulation of the INTEC calcine waste as it applies to transportation or disposal of calcine waste. Results relating to long-term disposal include: (1) a characterization of the pilot calcine waste; (2) Toxicity Characteristic Leaching Procedure (TCLP) testing of an optimum mixture of pilot calcine, polysiloxane and special additives; and, (3) Material Characterization Center testing MCC-1P evaluation of the optimum waste form. Results relating to transportation of the calcine material for a mixture of maximum waste loading include: compressive strength testing, 10-m drop test, melt testing, and a Department of Transportation (DOT) oxidizer test.

  15. CHARACTERIZATION OF DEFENSE NUCLEAR WASTE USING HAZARDOUS WASTE GUIDANCE. APPLICATIONS TO HANFORD SITE ACCELERATED HIGH-LEVEL WASTE TREATMENT AND DISPOSAL MISSION0

    SciTech Connect

    Hamel, William; Huffman, Lori; Lerchen, Megan; Wiemers, Karyn

    2003-02-27

    Federal hazardous waste regulations were developed for management of industrial waste. These same regulations are also applicable for much of the nation's defense nuclear wastes. At the U.S. Department of Energy's (DOE) Hanford Site in southeast Washington State, one of the nation's largest inventories of nuclear waste remains in storage in large underground tanks. The waste's regulatory designation and its composition and form constrain acceptable treatment and disposal options. Obtaining detailed knowledge of the tank waste composition presents a significant portion of the many challenges in meeting the regulatory-driven treatment and disposal requirements for this waste. Key in applying the hazardous waste regulations to defense nuclear wastes is defining the appropriate and achievable quality for waste feed characterization data and the supporting evidence demonstrating that applicable requirements have been met at the time of disposal. Application of a performance-based approach to demonstrating achievable quality standards will be discussed in the context of the accelerated high-level waste treatment and disposal mission at the Hanford Site.

  16. Impact of transporting defense high-level waste to a geologic repository

    SciTech Connect

    Joy, D.S.; Shappert, L.B.; Boyle, J.W.

    1984-08-01

    This transportation study assumes that defense high-level waste is stored in three locations (the Savannah River, Hanford, and Idaho Falls plants) and may be disposed of in (1) a commercial repository or (2) a defense-only repository, either of which could be located at one of the five candidate sites; also documented is a preliminary analysis of the costs and risks of transporting defense high-level waste from the three storage sites to the five potential candidate repository sites. 17 references, 4 figures, 27 tables.

  17. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    SciTech Connect

    Bullen, D.B.; Gdowski, G.E. ); Weiss, H. )

    1988-06-01

    Three copper-based alloys, CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni), are being considered along with three austenitic candidates as possible materials for fabrication of containers for disposal of high-level radioactive waste. The waste will include spent fuel assemblies from reactors as well as high-level reprocessing wastes in borosilicate glass and will be sent to the prospective repository at Yucca Mountain, Nevada, for disposal. The containers must maintain mechanical integrity for 50 yr after emplacement to allow for retrieval of waste during the preclosure phase of repository operation. Containment is required to be substantially complete for up to 300 to 1000 yr. During the early period, the containers will be exposed to high temperatures and high gamma radiation fields from the decay of high-level waste. The final closure joint will be critical to the integrity of the containers. This volume surveys the available data on the metallurgy of the copper-based candidate alloys and the welding techniques employed to join these materials. The focus of this volume is on the methods applicable to remote-handling procedures in a hot-cell environment with limited possibility of postweld heat treatment. The three copper-based candidates are ranked on the basis of the various closure techniques. On the basis of considerations regarding welding, the following ranking is proposed for the copper-based alloys: CDA 715 (best) > CDA 102 > CDA 613 (worst). 49 refs., 15 figs., 1 tab.

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

    SciTech Connect

    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.

  19. Silicon polymer encapsulation of high level calcine waste for transportation or disposal

    SciTech Connect

    G. G. Loomis

    2000-02-27

    Engineers at the Idaho National Engineering and Environmental Laboratory (INEEL) are investigating the use of a proprietary silicon-polymer to encapsulate high-level calcine waste stored at the INEEL's Idaho Nuclear Technology and Engineering Center (INTEC). The silicon-polymer-encapsulated waste may be suitable for direct disposal at a radioactive waste disposal facility or for transport to an offsite melter for further processing. In connection with silicon-polymer encapsulation, the University of Akron, under special arrangement with Orbit Technologies, the originator of the Polymer Encapsulation Technology (PET), has studied a simulated waste material from INTEC called pilot-scale calcine that contains hazardous materials but no radioactive isotopes. In this study, Toxicity Characteristic Leaching Procedure (TCLP) and Materials Characterization Center Test 1P were performed to test the waste form for disposal. In addition, a maximum waste loading was established for transporting the calcine waste at INTEC to an offsite melter. For this maximum waste loading, compressive strength testing, 10-m drop testing, melt testing, and a Department of Transportation (DOT) oxidizer test were performed.

  20. Silicon Polymer Encapsulation of High Level Calcine Waste for Transportation or Disposal

    SciTech Connect

    Loomis, Guy George; Miller, Carla Jean; Kimmel, Richard John

    2000-03-01

    Engineers at the Idaho National Engineering and Environmental Laboratory (INEEL) are investigating the use of a proprietary silicon-polymer to encapsulate high-level calcine waste stored at the INEEL's Idaho Nuclear Technology and Engineering Center (INTEC). The silicon-polymer-encapsulated waste may be suitable for direct disposal at a radioactive waste disposal facility or for transport to an offsite melter for further processing. In connection with silicon-polymer encapsulation, the University of Akron, under special arrangement with Orbit Technologies, the originator of the Polymer Encapsulation Technology (PET), has studied a simulated waste material from INTEC called pilot-scale calcine that contains hazardous materials but no radioactive isotopes. In this study, Toxicity Characteristic Leaching Procedure (TCLP) and Materials Characterization Center Test 1P were performed to test the waste form for disposal. In addition, a maximum waste loading was established for transporting the calcine waste at INTEC to an offsite melter. For this maximum waste loading, compressive strength testing, 10-m drop testing, melt testing, and a Department of Transportation (DOT) oxidizer test were performed.

  1. Natural geochemical analogues of the near field of high-level nuclear waste repositories

    SciTech Connect

    Apps, J.A.

    1995-09-01

    United States practice has been to design high-level nuclear waste (HLW) geological repositories with waste densities sufficiently high that repository temperatures surrounding the waste will exceed 100{degrees}C and could reach 250{degrees}C. Basalt and devitrified vitroclastic tuff are among the host rocks considered for waste emplacement. Near-field repository thermal behavior and chemical alteration in such rocks is expected to be similar to that observed in many geothermal systems. Therefore, the predictive modeling required for performance assessment studies of the near field could be validated and calibrated using geothermal systems as natural analogues. Examples are given which demonstrate the need for refinement of the thermodynamic databases used in geochemical modeling of near-field natural analogues and the extent to which present models can predict conditions in geothermal fields.

  2. Ceramic process and plant design for high-level nuclear waste immobilization

    SciTech Connect

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

    1983-01-01

    In the last 3 years, significant advances in ceramic technology for high-level nuclear waste solidification have been made. Product quality in terms of leach-resistance, compositional uniformity, structural integrity, and thermal stability promises to be superior to borosilicate glass. This paper addresses the process effectiveness and preliminary designs for glass and ceramic immobilization plants. The reference two-step ceramic process utilizes fluid-bed calcination (FBC) and hot isostatic press (HIP) consolidation. Full-scale demonstration of these well-developed processing steps has been established at DOE and/or commercial facilities for processing radioactive materials. Based on Savannah River-type waste, our model predicts that the capital and operating cost for the solidification of high-level nuclear waste is about the same for the ceramic and glass options. However, when repository costs are included, the ceramic option potentially offers significantly better economics due to its high waste loading and volume reduction. Volume reduction impacts several figures of merit in addition to cost such as system logistics, storage, transportation, and risk. The study concludes that the ceramic product/process has many potential advantages, and rapid deployment of the technology could be realized due to full-scale demonstrations of FBC and HIP technology in radioactive environments. Based on our finding and those of others, the ceramic innovation not only offers a viable backup to the glass reference process but promises to be a viable future option for new high-level nuclear waste management opportunities.

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

    SciTech Connect

    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.%.

  4. International program to study subseabed disposal of high-level radioactive wastes

    SciTech Connect

    Carlin, E.M.; Hinga, K.R.; Knauss, J.A.

    1984-01-01

    This report provides an overview of the international program to study seabed disposal of nuclear wastes. Its purpose is to inform legislators, other policy makers, and the general public as to the history of the program, technological requirements necessary for feasibility assessment, legal questions involved, international coordination of research, national policies, and research and development activities. Each of these major aspects of the program is presented in a separate section. The objective of seabed burial, similar to its continental counterparts, is to contain and to isolate the wastes. The subseabed option should not be confuesed with past practices of ocean dumping which have introduced wastes into ocean waters. Seabed disposal refers to the emplacement of solidified high-level radioactive waste (with or without reprocessing) in certain geologically stable sediments of the deep ocean floor. Specially designed surface ships would transport waste canisters from a port facility to the disposal site. Canisters would be buried from a few tens to a few hundreds of meters below the surface of ocean bottom sediments, and hence would not be in contact with the overlying ocean water. The concept is a multi-barrier approach for disposal. Barriers, including waste form, canister, ad deep ocean sediments, will separate wastes from the ocean environment. High-level wastes (HLW) would be stabilized by conversion into a leach-resistant solid form such as glass. This solid would be placed inside a metallic canister or other type of package which represents a second barrier. The deep ocean sediments, a third barrier, are discussed in the Feasibility Assessment section. The waste form and canister would provide a barrier for several hundred years, and the sediments would be relied upon as a barrier for thousands of years. 62 references, 3 figures, 2 tables.

  5. Preliminary Waste Form Compliance Plan for the Idaho National Engineering and Environmental Laboratory High-Level Waste

    SciTech Connect

    B. A. Staples; T. P. O'Holleran

    1999-05-01

    The Department of Energy (DOE) has specific technical and documentation requirements for high-level waste (HLW) that is to be placed in a federal repository. This document describes in general terms the strategy to be used at the Idaho National Engineering and Environmental Laboratory (INEEL) to demonstrate that vitrified HLW, if produced at the INEEL, meets these requirements. Waste form, canister, quality assurance, and documentation specifications are discussed. Compliance strategy is given, followed by an overview of how this strategy would be implemented for each specification.

  6. Lead-iron phosphate glass as a containment medium for the disposal of high-level nuclear wastes

    DOEpatents

    Boatner, L.A.; Sales, B.C.

    1984-04-11

    Disclosed are lead-iron phosphate glasses containing a high level of Fe/sub 2/O/sub 3/ for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste

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

    SciTech Connect

    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.

  8. INCREASING SAFETY AND REDUCING ENVIRONMENTAL DAMAGE RISK FROM AGING HIGH-LEVEL RADIOACTIVE WASTE TANKS

    EPA Science Inventory

    There exists a paramount need for improved understanding of the behavior of high-level nuclear waste containers and the impact on structural integrity in terms of leak tightness and mechanical stability. The current program, which at the time of this writing is in its early stage...

  9. CHEMICAL DECOMPOSITION OF HIGH-LEVEL NUCLEAR WASTE STORAGE/DISPOSAL GLASSES UNDER IRRADIATION

    EPA Science Inventory

    The Offices of Energy Research and Environmental Management are immediately concerned with the development of storage/immobilization media for high-level nuclear wastes and excess weapons plutonium. These media must be stable and free of risk to the public or to the environment f...

  10. Structural integrity and potential failure modes of hanford high-level waste tanks

    SciTech Connect

    Han, F.C.

    1996-09-30

    Structural Integrity of the Hanford High-Level Waste Tanks were evaluated based on the existing Design and Analysis Documents. All tank structures were found adequate for the normal operating and seismic loads. Potential failure modes of the tanks were assessed by engineering interpretation and extrapolation of the existing engineering documents.

  11. LIQUIDUS TEMPERATURE OF HIGH-LEVEL WASTE BOROSILICATE GLASSES WITH SPINEL PRIMARY PHASE

    EPA Science Inventory

    Liquidus temperatures (TL) were measured for high-level waste (HLW) borosilicate glasses covering a Savannah River composition region. The primary crystallization phase for most glasses was spinel, a solid solution of trevorite (NiFe2O4) with other oxides (FeO, MnO, and Cr2O3). T...

  12. FINAL REPORT. SETTLING OF SPINEL IN A HIGH-LEVEL WASTE GLASS MELTER

    EPA Science Inventory

    The research on spinel settling in the high-level waste (HLW) glass melter was conducted to assist HLW retrieval, glass formulation, feed preparation, and melter design and operation for minimum-risk and minimum-cost vitrification of HLW at Hanford and other DOE sites. The main r...

  13. PROGRESS REPORT. MECHANISMS AND KINETICS OF ORGANIC AGING IN HIGH-LEVEL WASTE

    EPA Science Inventory

    The objective is to develop predictive models of organic degradation in high-level wastes (HLW). We make this information available to facility operators on the Hanford Site to support decision-making processes regarding safety, retrieval, and treatment issues. Emphasis is placed...

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

    SciTech Connect

    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.

  15. Design of equipment used for high-level waste vitrification at the West Valley Demonstration Project

    SciTech Connect

    Vance, R.F.; Brill, B.A.; Carl, D.E.

    1997-06-01

    The equipment as designed, started, and operated for high-level radioactive waste vitrification at the West Valley Demonstration Project in western New York State is described. Equipment for the processes of melter feed make-up, vitrification, canister handling, and off-gas treatment are included. For each item of equipment the functional requirements, process description, and hardware descriptions are presented.

  16. Heat transfer effects in vertically emplaced high level nuclear waste container

    SciTech Connect

    Moujaes, S.F.; Lei, Y.M.

    1994-05-01

    Modeling free convection heat transfer in an cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the high level nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rack. These waste containers are vertically emplaced in the borehole 300 meters below ground, and in a horizontal grid of 30 {times} 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3--4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions.

  17. Corrosion Management of the Hanford High-Level Nuclear Waste Tanks

    NASA Astrophysics Data System (ADS)

    Beavers, John A.; Sridhar, Narasi; Boomer, Kayle D.

    2014-03-01

    The Hanford site is located in southeastern Washington State and stores more than 200,000 m3 (55 million gallons) of high-level radioactive waste resulting from the production and processing of plutonium. The waste is stored in large carbon steel tanks that were constructed between 1943 and 1986. The leak and structurally integrity of the more recently constructed double-shell tanks must be maintained until the waste can be removed from the tanks and encapsulated in glass logs for final disposal in a repository. There are a number of corrosion-related threats to the waste tanks, including stress-corrosion cracking, pitting corrosion, and corrosion at the liquid-air interface and in the vapor space. This article summarizes the corrosion management program at Hanford to mitigate these threats.

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

    SciTech Connect

    Crawford, T.W.

    1995-09-22

    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.

  19. Operating experience during high-level waste vitrification at the West Valley Demonstration Project

    SciTech Connect

    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.

  20. National long-term high-level waste-technology program

    NASA Astrophysics Data System (ADS)

    Gray, P. L.

    The national program for long-term management of high level waste (HLW) from nuclear fuels reprocessing is discussed. This covers only DOE defense wastes. Current emphasis is on solidification of waste into a form that, along with additional barriers, may be permanently stored in a repository. An integrated national plan incorporates all the elements of such an overall HLW disposal system. Interim storage is in near-surface tanks at the Hanford and Savannah River sites. At the Idaho site, waste is stored in bins after being calcined. Some Idaho waste is liquid and is also stored in tanks before calcination. Retrieval and immobilization of HLW into a solid, low-release form represent the major elements for which our long-term program has responsibility. Once solidified, the waste will temporarily remain onsite until the final disposal site is prepared for receipt of waste. Currently, a geologic repository is favored for ultimate disposal, although other possibilities such as seabed, icecap, space, and near-surface disposal are also being considered.

  1. Four themes that underlie the high-level nuclear waste management program

    SciTech Connect

    Sprecher, W.M.

    1989-01-01

    In 1982, after years of deliberation and in response to mounting pressures from environmental, industrial, and other groups, the US Congress enacted the Nuclear Waste Policy Act (NWPA) of 1982, which was signed into law by the President in January 1983. That legislation signified a major milestone in the nation's management of high-level nuclear waste, since it represented a consensus among the nation's lawmakers to tackle a problem that had evaded solution for decades. Implementation of the NWPA has proven to be exceedingly difficult, as attested by the discord generated by the US Department of Energy's (DOE's) geologic repository and monitored retrievable storage (MRS) facility siting activities. The vision that motivated the crafters of the 1982 act became blurred as opposition to the law increased. After many hearings that underscored the public's concern with the waste management program, the Congress enacted the Nuclear Waste Policy Amendments Act of 1987 (Amendments Act), which steamlined and focused the program, while establishing three independent bodies: the MRS Review Commission, the Nuclear Waste Technical Review Board, and the Office of the Nuclear Waste Negotiator. Yet, even as the program evolves, several themes characterizing the nation's effort to solve the waste management problem continue to prevail. The first of these themes has to do with social consciousness, and the others that follow deal with technical leadership, public involvement and risk perceptions, and program conservatism.

  2. Selection of containment systems for commercial high-level radioactive waste management

    SciTech Connect

    Kaplan, M. F.; Giuffre, M. S.; Bartlett, J. W.

    1981-05-01

    This document reports the results of a study aimed at determining the best strategy for providing containment during management of commercial high-level radioactive wastes. Containment to assure public and worker safety is needed for all storage, transport, handling, and disposal operations. There are several thousand containment system options; this work determined, in overview rather than detail, which options should be pursued. This work shows that the geologic and engineered barriers in repositories in different geologic media, such as salt and granite, play very different roles in preserving long-term containment. In sum, there is no common engineered waste package that is suitable for disposal in all geologic media, each package must be tailored to the specific repository system. The need to make the waste package specific to the repository system leads to the key elements of waste management containment strategy: perform final packaging at the disposal site, and deliver to the site a waste that is in a form suitable for disposal and in a container that is (a) appropriate for the process that produced the waste form, (b) satisfactory for transport, and (c) suitable as the common basis for custom tailoring the waste package for any repository system. As described in this report, mild carbon steel is a container material that can be expected to meet these requirements.

  3. Reference design and operations for deep borehole disposal of high-level radioactive waste.

    SciTech Connect

    Herrick, Courtney Grant; Brady, Patrick Vane; Pye, Steven; Arnold, Bill Walter; Finger, John Travis; Bauer, Stephen J.

    2011-10-01

    A reference design and operational procedures for the disposal of high-level radioactive waste in deep boreholes have been developed and documented. The design and operations are feasible with currently available technology and meet existing safety and anticipated regulatory requirements. Objectives of the reference design include providing a baseline for more detailed technical analyses of system performance and serving as a basis for comparing design alternatives. Numerous factors suggest that deep borehole disposal of high-level radioactive waste is inherently safe. Several lines of evidence indicate that groundwater at depths of several kilometers in continental crystalline basement rocks has long residence times and low velocity. High salinity fluids have limited potential for vertical flow because of density stratification and prevent colloidal transport of radionuclides. Geochemically reducing conditions in the deep subsurface limit the solubility and enhance the retardation of key radionuclides. A non-technical advantage that the deep borehole concept may offer over a repository concept is that of facilitating incremental construction and loading at multiple perhaps regional locations. The disposal borehole would be drilled to a depth of 5,000 m using a telescoping design and would be logged and tested prior to waste emplacement. Waste canisters would be constructed of carbon steel, sealed by welds, and connected into canister strings with high-strength connections. Waste canister strings of about 200 m length would be emplaced in the lower 2,000 m of the fully cased borehole and be separated by bridge and cement plugs. Sealing of the upper part of the borehole would be done with a series of compacted bentonite seals, cement plugs, cement seals, cement plus crushed rock backfill, and bridge plugs. Elements of the reference design meet technical requirements defined in the study. Testing and operational safety assurance requirements are also defined. Overall

  4. Separation of americium, curium, and rare earths from high-level wastes by oxalate precipitation: experiments with synthetic waste solutions

    SciTech Connect

    Forsberg, C.W.

    1980-01-01

    The separation of trivalent actinides and rare earths from other fission products in high-level nuclear wastes by oxalate precipitation followed by ion exchange (OPIX) was experimentally investigated using synthetic wastes and a small-scale, continuous-flow oxalic acid precipitation and solid-liquid separation system. Trivalent actinide and rare earth oxalates are relatively insoluble in 0.5 to 1.0 M HNO/sub 3/ whereas other fission product oxalates are not. The continuous-flow system consisted of one or two stirred-tank reactors in series for crystal growth. Oxalic acid and waste solutions were mixed in the first tank, with the product solid-liquid slurry leaving the second tank. Solid-liquid separation was tested by filters and by a gravity settler. The experiments determined the fraction of rare earths precipitated and separated from synthetic waste streams as a function of number of reactors, system temperature, oxalic acid concentration, liquid residence time in the process, power input to the stirred-tank reactors, and method of solid-liquid separation. The crystalline precipitate was characterized with respect to form, size, and chemical composition. These experiments are only the first step in converting a proposed chemical flowsheet into a process flowsheet suitable for large-scale remote operations at high activity levels.

  5. Settling of Spinel in a High-Level Waste Glass Melter

    SciTech Connect

    Hrma, Pavel R.; Schill, Pert; Nemec, Lubomir

    2002-01-18

    High-level nuclear waste is being vitrified, i.e., converted to a durable glass that can be stored in a safe repository for hundreds of thousands of years. Waste vitrification is accomplished in reactors called melters to which the waste is charged together with glass-forming additives. The mixture is electrically heated to a temperature as high as 1150?C (or even higher in advanced melters) to create a melt that becomes glass on cooling. This process is slow and expensive. Moreover, the melters that are currently in use or are going to be used in the U.S. are sensitive to clogging and thus cannot process melt in which solid particles are suspended. These particles settle and gradually accumulate on the melter bottom. Such particles, most often small crystals of spinel (a mineral containing iron, nickel, chromium, and other minor oxides), inevitably occur in the melt when the content of the waste in the glass (called waste loading) increases above a certain limit. To avoid the presence of solid particles in the melter, the waste loading is kept rather low, in average 15% lower than in glass formulated for more robust melters.

  6. Settling of Spinel in A High-Level Waste Glass Melter

    SciTech Connect

    Pavel Hrma; Pert Schill; Lubomir Nemec

    2002-01-07

    High-level nuclear waste is being vitrified, i.e., converted to a durable glass that can be stored in a safe repository for hundreds of thousands of years. Waste vitrification is accomplished in reactors call melters to which the waste is charged together with glass-forming additives. The mixture is electrically heated to a temperature as high as 1150 degree C (or even higher in advanced melters) to create a melt that becomes glass on cooling. This process is slow and expensive. Moreover, the melters that are currently in use or are going to be used in the U.S. are sensitive to clogging and thus cannot process melt in which solid particles are suspended. These particles settle and gradually accumulate on the melter bottom. Such particles, most often small crystals of spinel ( a mineral containing iron, nickel, chromium, and other minor oxides), inevitably occurred in the melt when the content of the waste in the glass (called waste loading) increases above a certain limit. To avoid the presence of solid particles in the melter, the waste loading is kept rather low, in average 15% lower than in glass formulated for more robust melters.

  7. Development, evaluation, and selection of candidate high-level waste forms

    SciTech Connect

    Bernadzikowski, T A; Allender, J S; Gordon, D E; Gould, Jr, T H

    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.

  8. Alternative-waste-form evaluation for Savannah River Plant high-level waste

    SciTech Connect

    Gould, Jr, T H; Crandall, J L

    1982-01-01

    Results of the waste form evaluation are summarized as: risks of human exposure are comparable and extremely small for either borosilicate glass or Synroc ceramic. Waste form properties are more than adequate for either form. The waste form decision can therefore be made on the basis of practicality and cost effectiveness. Synroc offers lower costs for transportation and emplacement. The borosilicate glass form offers the lowest total disposal cost, much simpler and less costly production, an established and proven process, lower future development costs, and an earlier startup of the DWPF.

  9. Validation of Stress Corrosion Cracking Model for High Level Radioactive-Waste Packages

    SciTech Connect

    Lu, S; Gordon, G; Andresen, P

    2004-04-22

    A stress corrosion cracking (SCC) model has been adapted for performance prediction of high level radioactive-waste packages to be emplaced in the proposed Yucca Mountain radioactive-waste repository. SCC is one form of environmentally assisted cracking resulting from the presence of three factors: metallurgical susceptibility, critical environment, and tensile stresses. For waste packages of the proposed Yucca Mountain repository, the outer barrier material is the highly corrosion-resistant Alloy UNS-N06022, the environment is represented by the water film present on the surface of the waste package from dripping or deliquescence of soluble salts present in any surface deposits, and the stress is principally the weld induced residual stress. SCC has historically been separated into 'initiation' and 'propagation' phases. Initiation of SCC will not occur on a smooth surface if the surface stress is below a threshold value defined as the threshold stress. Cracks can also initiate at and propagate from flaws (or defects) resulting from manufacturing processes (such as welding). To account for crack propagation, the slip dissolution/film rupture (SDFR) model is adopted to provide mathematical formulae for prediction of the crack growth rate. Once the crack growth rate at an initiated SCC is determined, it can be used by the performance assessment (not in the scope of this paper) to determine the time to through-wall penetration for the waste package. This paper presents the development and validation of the SDFR crack growth rate model based on technical information in the literature as well as experimentally determined crack growth rates developed specifically for Alloy UNS- N06022 in environments relevant to high level radioactive-waste packages of the proposed Yucca Mountain radioactive-waste repository.

  10. Annotated bibliography for the design of waste packages for geologic disposal of spent fuel and high-level waste

    SciTech Connect

    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.