WorldWideScience

Sample records for high-level waste storage

  1. Handling and storage of conditioned high-level wastes

    International Nuclear Information System (INIS)

    Heafield, W.

    1984-01-01

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

  2. Handling and storage of conditioned high-level wastes

    International Nuclear Information System (INIS)

    1983-01-01

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

  3. High-level radioactive waste glass and storage canister design

    International Nuclear Information System (INIS)

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

    1979-01-01

    Management of high-level radioactive wastes is a primary concern in nuclear operations today. The main objective in managing these wastes is to convert them into a solid, durable form which is then isolated from man. A description is given of the design and evaluation of this waste form. The waste form has two main components: the solidified waste and the storage canister. The solid waste form discussed in this study is glass. Waste glasses have been designed to be inert to water attack, physically rugged, low in volatility, and stable over time. Two glass-making processes are under development at PNL. The storage canister is being designed to provide high-integrity containment for solidified wastes from processing to terminal storage. An outline is given of the steps in canister design: material selection, stress and thermal analyses, quality verification, and postfill processing. Examples are given of results obtained from actual nonradioactive demonstration tests. 14 refs

  4. Spent fuel and high-level radioactive waste storage

    International Nuclear Information System (INIS)

    Trigerman, S.

    1988-06-01

    The subject of spent fuel and high-level radioactive waste storage, is bibliographically reviewed. The review shows that in the majority of the countries, spent fuels and high-level radioactive wastes are planned to be stored for tens of years. Sites for final disposal of high-level radioactive wastes have not yet been found. A first final disposal facility is expected to come into operation in the United States of America by the year 2010. Other final disposal facilities are expected to come into operation in Germany, Sweden, Switzerland and Japan by the year 2020. Meanwhile , stress is placed upon the 'dry storage' method which is carried out successfully in a number of countries (Britain and France). In the United States of America spent fuels are stored in water pools while the 'dry storage' method is still being investigated. (Author)

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

    International Nuclear Information System (INIS)

    Deffenbaugh, M.L.

    1997-01-01

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

  6. Thermo-aeraulics of high level waste storage facilities

    International Nuclear Information System (INIS)

    Lagrave, Herve; Gaillard, Jean-Philippe; Laurent, Franck; Ranc, Guillaume; Duret, Bernard

    2006-01-01

    This paper discusses the research undertaken in response to axis 3 of the 1991 radioactive waste management act, and possible solutions concerning the processes under consideration for conditioning and long-term interim storage of long-lived radioactive waste. The notion of 'long-term' is evaluated with respect to the usual operating lifetime of a basic nuclear installation, about 50 years. In this context, 'long-term' is defined on a secular time scale: the lifetime of the facility could be as long as 300 years. The waste package taken into account is characterized notably by its high thermal power release. Studies were carried out in dedicated facilities for vitrified waste and for spent UOX and MOX fuel. The latter are not considered as wastes, owing to the value of the reusable material they contain. Three primary objectives have guided the design of these long-term interim storage facilities: - ensure radionuclide containment at all times; - permit retrieval of the containers at any time; - minimize surveillance; - maintenance costs. The CEA has also investigated surface and subsurface facilities. It was decided to work on generic sites with a reasonable set of parameters values that should be applicable at most sites in France. All the studies and demonstrations to date lead to the conclusion that long-term interim storage is technically feasible. The paper addresses the following items: - Long-term interim storage concepts for high-level waste; - Design principles and options for the interim storage facilities; - General architecture; - Research topics, Storage facility ventilation, Dimensioning of the facility; - Thermo-aeraulics of a surface interim storage facility; - VALIDA surface loop, VALIDA single container test campaign, Continuation of the VALIDA program; - Thermo-aeraulics of a network of subsurface interim storage galleries; - SIGAL subsurface loop; - PROMETHEE subsurface loop; - Temperature behaviour of the concrete structures; - GALATEE

  7. Thermo-aeraulics of high level waste storage facilities

    Energy Technology Data Exchange (ETDEWEB)

    Lagrave, Herve; Gaillard, Jean-Philippe; Laurent, Franck; Ranc, Guillaume [CEA/Valrho, B.P. 17171, F-30207 Bagnols-sur-Ceze (France); Duret, Bernard [CEA Grenoble, 17 rue des Martyrs, 38054 Grenoble cedex 9 (France)

    2006-07-01

    This paper discusses the research undertaken in response to axis 3 of the 1991 radioactive waste management act, and possible solutions concerning the processes under consideration for conditioning and long-term interim storage of long-lived radioactive waste. The notion of 'long-term' is evaluated with respect to the usual operating lifetime of a basic nuclear installation, about 50 years. In this context, 'long-term' is defined on a secular time scale: the lifetime of the facility could be as long as 300 years. The waste package taken into account is characterized notably by its high thermal power release. Studies were carried out in dedicated facilities for vitrified waste and for spent UOX and MOX fuel. The latter are not considered as wastes, owing to the value of the reusable material they contain. Three primary objectives have guided the design of these long-term interim storage facilities: - ensure radionuclide containment at all times; - permit retrieval of the containers at any time; - minimize surveillance; - maintenance costs. The CEA has also investigated surface and subsurface facilities. It was decided to work on generic sites with a reasonable set of parameters values that should be applicable at most sites in France. All the studies and demonstrations to date lead to the conclusion that long-term interim storage is technically feasible. The paper addresses the following items: - Long-term interim storage concepts for high-level waste; - Design principles and options for the interim storage facilities; - General architecture; - Research topics, Storage facility ventilation, Dimensioning of the facility; - Thermo-aeraulics of a surface interim storage facility; - VALIDA surface loop, VALIDA single container test campaign, Continuation of the VALIDA program; - Thermo-aeraulics of a network of subsurface interim storage galleries; - SIGAL subsurface loop; - PROMETHEE subsurface loop; - Temperature behaviour of the concrete

  8. Storage of High Level Nuclear Waste in Germany

    Directory of Open Access Journals (Sweden)

    Dietmar P. F. Möller

    2007-01-01

    Full Text Available Nuclear energy is very often used to generate electricity. But first the energy must be released from atoms what can be done in two ways: nuclear fusion and nuclear fission. Nuclear power plants use nuclear fission to produce electrical energy. The electrical energy generated in nuclear power plants does not produce polluting combustion gases but a renewable energy, an important fact that could play a key role helping to reduce global greenhouse gas emissions and tackling global warming especially as the electricity energy demand rises in the years ahead. This could be assumed as an ideal win-win situation, but the reverse site of the medal is that the production of high-level nuclear waste outweighs this advantage. Hence the paper attempt to highlight the possible state-of-art concepts for the safe and sustaining storage of high-level nuclear waste in Germany.

  9. Glasses used for the high level radioactive wastes storage

    International Nuclear Information System (INIS)

    Sombret, C.

    1983-06-01

    High level radioactive wastes generated by the reprocessing of spent fuels is an important concern in the conditioning of radioactive wastes. This paper deals with the status of the knowledge about glasses used for the treatment of these liquids [fr

  10. Retrievable surface storage: interim storage of solidified high-level waste

    International Nuclear Information System (INIS)

    LaRiviere, J.R.; Nelson, D.C.

    1976-01-01

    Studies have been conducted on retrievable-surface-storage concepts for the interim storage of solidified high-level wastes. These studies have been reviewed by the Panel on Engineered Storage, convened by the Committee on Radioactive Waste Management of the National Research Council-National Academy of Sciences. The Panel has concluded that ''retrievable surface storage is an acceptable interim stage in a comprehensive system for managing high-level radioactive wastes.'' The scaled storage cask concept, which was recommended by the Panel on Engineered Storage, consists of placing a canister of waste inside a carbon-steel cask, which in turn is placed inside a thick concrete cylinder. The waste is cooled by natural convection air flow through an annulus between the cask and the inner wall of the concrete cylinder. The complete assembly is placed above ground in an outdoor storage area

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

    Science.gov (United States)

    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)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  13. Laboratory simulation of high-level liquid waste evaporation and storage

    International Nuclear Information System (INIS)

    Anderson, P.A.

    1978-01-01

    The reprocessing of nuclear fuel generates high-level liquid wastes (HLLW) which require interim storage pending solidification. Interim storage facilities are most efficient if the HLLW is evaporated prior to or during the storage period. Laboratory evaporation and storage studies with simulated waste slurries have yielded data which are applicable to the efficient design and economical operation of actual process equipment

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

    International Nuclear Information System (INIS)

    CALMUS, R.B.

    1999-01-01

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

  15. System Specification for Immobilized High-Level Waste Interim Storage

    International Nuclear Information System (INIS)

    CALMUS, R.B.

    2000-01-01

    This specification establishes the system-level functional, performance, design, interface, and test requirements for Phase 1 of the IHLW Interim Storage System, located at the Hanford Site in Washington State. The IHLW canisters will be produced at the Hanford Site by a Selected DOE contractor. Subsequent to storage the canisters will be shipped to a federal geologic repository

  16. Pre-disposal storage, transport and handling of vitrified high level waste

    International Nuclear Information System (INIS)

    Kempe, T.F.; Martin, A.

    1981-05-01

    The objectives of the study were to review non site-specific engineering features of the storage, transport and handling of vitrified high level radioactive waste prior to its transfer into an underground repository, and to identify those features which require validation or development. Section headings are: introduction (historical and technical background); characteristics and arisings of vitrified high level waste; overpacks (additional containment barrier, corrosion resistant); interim storage of HLW; transport of HLW; handling; conclusions and recommendations. (U.K.)

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

    Energy Technology Data Exchange (ETDEWEB)

    ROGERS, C.A.

    2000-02-17

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

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

    International Nuclear Information System (INIS)

    ROGERS, C.A.

    2000-01-01

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

  19. Handling and storage of high-level radioactive liquid wastes requiring cooling

    International Nuclear Information System (INIS)

    1979-01-01

    The technology of high-level liquid wastes storage and experience in this field gained over the past 25 years are reviewed in this report. It considers the design requirements for storage facilities, describes the systems currently in use, together with essential accessories such as the transfer and off-gas cleaning systems, and examines the safety and environmental factors

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

    International Nuclear Information System (INIS)

    Burgard, K.C.

    1998-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Burgard, K.C.

    1998-04-09

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

  2. Deep geologic storage of high level radioactive wastes: conceptual generic designs

    International Nuclear Information System (INIS)

    1995-01-01

    This report summarizes the studies on deep geologic storage of radioactive wastes and specially for the high-level radioactive wastes. The study is focussed to the geotechnical assessment and generic-conceptual designs. Methodology analysis, geotechnical feasibility, costs and operation are studied

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

    International Nuclear Information System (INIS)

    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

  4. Near-surface storage facilities for vitrified high-level wastes

    International Nuclear Information System (INIS)

    Kondrat'ev, A.N.; Kulichenko, V.V.; Kryukov, I.I.; Krylova, N.V.; Paramoshkin, V.I.; Strakhov, M.V.

    1980-01-01

    Concurrently with the development of methods for solidifying liquid radioactive wastes, reliable and safe methods for the storage and disposal of solidified wastes are being devised in the USSR and other countries. One of the main factors affecting the choice of storage conditions for solidified wastes originating from the vitrification of high-level liquid wastes from fuel reprocessing plants is the problem of removing the heat produced by radioactive decay. In order to prevent the temperature of solidified wastes from exceeding the maximum permissible level for the material concerned, it is necessary to limit either the capacity of waste containers or the specific heat release of the wastes themselves. In order that disposal of high-level wastes in geological formations should be reliable and economic, solidified wastes undergo interim storage in near-surface storage facilities with engineered cooling systems. The paper demonstrates the relative influences of specific heat release, of the maximum permissible storage temperature for vitrified wastes and of the methods chosen for cooling wastes in order for the dimensions of waste containers to be reduced to the extent required. The effect of concentrating wastes to a given level in the vitrification process on the cost of storage in different types of storage facility is also examined. Calculations were performed for the amount of vitrified wastes produced by a reprocessing plant with a capacity of five tonnes of uranium per 24 hours. Fuel elements from reactors of the water-cooled, water-moderated type are sent for reprocessing after having been held for about two years. The dimensions of the storage facility are calculated on the assumption that it will take five years to fill

  5. Numerical simulation on stir system of jet ballast in high level liquid waste storage tank

    International Nuclear Information System (INIS)

    Lu Yingchun

    2012-01-01

    The stir system of jet ballast in high level liquid waste storage tank was simulation object. Gas, liquid and solid were air, sodium nitrate liquor and titanium whitening, respectively. The mathematic model based on three-fluid model and the kinetic theory of particles was established for the stir system of jet ballast in high level liquid waste storage tank. The CFD commercial software was used for solving this model. The detail flow parameters as three phase velocity, pressure and phase loadings were gained. The calculated results agree with the experimental results, so they can well define the flow behavior in the tank. And this offers a basic method for the scale-up and optimization design of the stir system of jet ballast in high level liquid waste storage tank. (author)

  6. Plan of deep underground construction for investigations on high-level radioactive waste storage

    International Nuclear Information System (INIS)

    Mayanovskij, M.S.

    1996-01-01

    The program of studies of the Japanese PNC corporation on construction of deep underground storage for high-level radioactive wastes is presented. The program is intended for 20 years. The total construction costs equal about 20 billion yen. The total cost of the project is equal to 60 billion yen. The underground part is planned to reach 1000 m depth

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  8. Storage and disposal of high-level radioactive waste from advanced FBR fuel cycle

    International Nuclear Information System (INIS)

    Nishihara, Kenji; Oigawa, Hiroyuki; Nakayama, Shinichi; Ono, Kiyoshi; Shiotani, Hiroki

    2011-01-01

    Waste management of fast breeder reactor (FBR) fuel cycle with and without partitioning and transmutation (P and T) technology was investigated by focusing on thermal constraints due to heat deposition from waste in storage and disposal facilities including economics aspects of those facilities. Partitioning of minor actinides (MAs) and heat-generating fission products in high-level waste can enlarge the containment ratio of waste elements in the glass waste forms and shorten predisposal storage period. Though MAs can be transmuted in FBRs or dedicated transmuters, heat-generating fission products are difficult to be transmuted; they are partitioned and stored for a long time before disposal. The disposal concepts for heat-generating fission products and remainders such as rare-earth elements depend on storage period that ranges from several years to several hundreds of years. Short-term storage results in small size of storage facilities and large size of repositories, and vice versa for long-term storage. This trade-off relation was analyzed by estimating repository size as a function of storage period. The result shows that transmutation of MAs is essentially effective to reduce repository size regardless to storage period, and a combination of P and T can provide a smaller repository than the conventional one by two orders of magnitude. The cost analysis for waste management was also made based on rough assumptions on storage, transportation and repository excluding cost for introducing P and T that are still under evaluation. Cost of waste management for FBR without P and T is 0.25 Yen/kWh that is slightly smaller than that for LWR without P and T, 0.30 Yen/kWh. The introduction of MA transmutation to the FBR results in cost of 0.20 Yen/kWh, and full introduction of P and T provides the smallest cost of 0.08 Yen/kWh. (author)

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

    Science.gov (United States)

    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.

  10. Treatment and Storage of High-Level Radioactive Wastes. Proceedings of the Symposium on Treatment and Storage of High-Level Radioactive Wastes

    International Nuclear Information System (INIS)

    1963-01-01

    A variety of radioactive materials having no immediate use result from the utilization of atomic energy. The manner in which these materials are handled has repercussions on reactor economy and technology, on the health and safety of persons and populations and on atomic legislation. Excellent progress has been made in developing a technology capable of safely and economically dealing with these materials so that no immediate problems exist. The highly radioactive ''wastes'' arising from the reprocessing of irradiated fuel pose long-range problems, however, and methods for the ultimate disposal of these wastes must be developed and evaluated. Such development and evaluation can be materially assisted by providing the scientists doing the work with an opportunity of exchanging ideas and information on their experience. Therefore, the IAEA, as part of its programme of promoting nuclear technology, convened in Vienna from 8-12 October 1962 the Symposium on the Treatment and Storage of High-level Radioactive Wastes. The Symposium was attended by 130 scientists from 19 countries and two international organizations. Thirty-three papers were presented and discussed in full and formed a background for a panel discussion of chairmen near the end of the Symposium. The papers and a record of the discussions are published in this single volume. It is hoped that the information thus recorded will achieve the desired purpose of assisting the peaceful development of atomic energy

  11. Mineral-modeled ceramics for long-term storage of high-level nuclear wastes

    International Nuclear Information System (INIS)

    Vance, E.R.

    1980-01-01

    Over the past ten years, Penn State's Materials Research Laboratory has done extensive work on mineral-modeled ceramics for high-level nuclear waste storage. These ceramics are composed of several mineral analogues that form a monolithic polycrystalline aggregate. Mineral-modeling can be made in a similar fashion to nuclear waste glasses, and their naturally occurring analogues are known to last millions, and even billions, of years in hot, wet conditions. It is believed that such ceramics could reduce dispersal of radionuclides by leaching to a minimum

  12. High level nuclear wastes

    International Nuclear Information System (INIS)

    Lopez Perez, B.

    1987-01-01

    The transformations involved in the nuclear fuels during the burn-up at the power nuclear reactors for burn-up levels of 33.000 MWd/th are considered. Graphs and data on the radioactivity variation with the cooling time and heat power of the irradiated fuel are presented. Likewise, the cycle of the fuel in light water reactors is presented and the alternatives for the nuclear waste management are discussed. A brief description of the management of the spent fuel as a high level nuclear waste is shown, explaining the reprocessing and giving data about the fission products and their radioactivities, which must be considered on the vitrification processes. On the final storage of the nuclear waste into depth geological burials, both alternatives are coincident. The countries supporting the reprocessing are indicated and the Spanish programm defined in the Plan Energetico Nacional (PEN) is shortly reviewed. (author) 8 figs., 4 tabs

  13. Intermediate storage facility for vitrified high level waste from the reprocessing of spent nuclear fuel

    International Nuclear Information System (INIS)

    1978-04-01

    An intermediate storage facility for vitrified high level waste is described. The design was made specifically for Swedish conditions but can due to modular design be applied also for other conditions. Most of the plant is located underground with a rock cover of about 30 m in order to provide protection against external forces such as acts of war and sabotage. The storage area consists of four caverns each with 150 pits. Each pit can take 10 waste cylinders of 0.4 m diameter and 1.5 m length containing 150 liters of glass. The capacity can be increased by adding additional caverns. Cooling is obtained by forced air convection. Reception areas, auxiliary systems and operation of the plant are also described

  14. An overview of the DOE high-level waste storage tank structural integrity assessment guidelines

    International Nuclear Information System (INIS)

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

    1995-01-01

    The basic elements of a structural integrity program for high-level waste storage tanks include identifying significant aging degradation mechanisms, developing programs to monitor and control these degradation processes, and developing management options and procedures to minimize impact on the environment should tank leakage develop. A Waste Tank Structural Integrity Panel (TSIP) was established by Brookhaven National Laboratory at the request of the DOE Office of Environmental Restoration and Waste Management to review these elements and prepare a set of guidelines that could be used by DOE and its contractors to manage the structural integrity of these tanks. These guidelines emphasize the identification of significant degradation mechanisms for both the steel and concrete components of the tanks, the recommended monitoring and inspection programs, and the indicated management options

  15. On-site storage of high level nuclear waste: Attitudes and perceptions of local residents

    International Nuclear Information System (INIS)

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

    1996-01-01

    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 countries 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. 24 refs., 7 figs., 5 tabs

  16. The storage of liquid high level waste at BNFL, Sellafield. Addendum to February 2000 report

    International Nuclear Information System (INIS)

    2001-08-01

    On 18 February 2000 the Health and Safety Executive (HSE) published a report on the work of its Nuclear Installations Inspectorate (NIl) in regulating the storage of liquid high level waste at the BNFL Sellafield site. Within the report NIl gave two undertakings. One was to publish an addendum around 1 year later covering its assessment of the new safety case for the storage plant and the second was to publish a further addendum when progress had been made with options studies for reducing the stocks of liquid high level waste (HLW), also referred to as highly active liquor (HAL), to a buffer level. A progress report was published in February 2001 which included a summary of the assessment of the new safety case and NIl's regulatory action to enforce liquid HLW stock reductions. This addendum provides a more detailed update on the position reached based on consideration of BNFL's responses to the recommendations from the February 2000 HLW report since its publication. It embodies the two addenda referred to above integrated into a single document for publication

  17. The management of high-level radioactive wastes

    International Nuclear Information System (INIS)

    Lennemann, Wm.L.

    1979-01-01

    The definition of high-level radioactive wastes is given. The following aspects of high-level radioactive wastes' management are discussed: fuel reprocessing and high-level waste; storage of high-level liquid waste; solidification of high-level waste; interim storage of solidified high-level waste; disposal of high-level waste; disposal of irradiated fuel elements as a waste

  18. Risk perception on management of nuclear high-level and transuranic waste storage

    Energy Technology Data Exchange (ETDEWEB)

    Dees, Lawrence A. [Colorado Christian Univ., Lakewood, CO (United States)

    1994-08-15

    The Department of Energy`s program for disposing of nuclear High-Level Waste (HLW) and transuranic (TRU) waste has been impeded by overwhelming political opposition fueled by public perceptions of actual risk. Analysis of these perceptions shows them to be deeply rooted in images of fear and dread that have been present since the discovery of radioactivity. The development and use of nuclear weapons linked these images to reality and the mishandling of radioactive waste from the nations military weapons facilities has contributed toward creating a state of distrust that cannot be erased quickly or easily. In addition, the analysis indicates that even the highly educated technical community is not well informed on the latest technology involved with nuclear HLW and TRU waste disposal. It is not surprising then, that the general public feels uncomfortable with DOE`s management plans for with nuclear HLW and TRU waste disposal. Postponing the permanent geologic repository and use of Monitored Retrievable Storage (MRS) would provide the time necessary for difficult social and political issues to be resolved. It would also allow time for the public to become better educated if DOE chooses to become proactive.

  19. Risk perception on management of nuclear high-level and transuranic waste storage

    International Nuclear Information System (INIS)

    Dees, L.A.

    1994-01-01

    The Department of Energy's program for disposing of nuclear High-Level Waste (HLW) and transuranic (TRU) waste has been impeded by overwhelming political opposition fueled by public perceptions of actual risk. Analysis of these perceptions shows them to be deeply rooted in images of fear and dread that have been present since the discovery of radioactivity. The development and use of nuclear weapons linked these images to reality and the mishandling of radioactive waste from the nations military weapons facilities has contributed toward creating a state of distrust that cannot be erased quickly or easily. In addition, the analysis indicates that even the highly educated technical community is not well informed on the latest technology involved with nuclear HLW and TRU waste disposal. It is not surprising then, that the general public feels uncomfortable with DOE's management plans for with nuclear HLW and TRU waste disposal. Postponing the permanent geologic repository and use of Monitored Retrievable Storage (MRS) would provide the time necessary for difficult social and political issues to be resolved. It would also allow time for the public to become better educated if DOE chooses to become proactive

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

    International Nuclear Information System (INIS)

    Wodrich, D.D.

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  2. Immobilized High-Level Waste (HLW) Interim Storage Alternative Generation and analysis and Decision Report - second Generation Implementing Architecture

    International Nuclear Information System (INIS)

    CALMUS, R.B.

    2000-01-01

    Two alternative approaches were previously identified to provide second-generation interim storage of Immobilized High-Level Waste (IHLW). One approach was retrofit modification of the Fuel and Materials Examination Facility (FMEF) to accommodate IHLW. The results of the evaluation of the FMEF as the second-generation IHLW interim storage facility and subsequent decision process are provided in this document

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

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

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

    International Nuclear Information System (INIS)

    1994-04-01

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

  5. Final storage high-level radioactive waste in Sweden - the way to the 2009 siting decision

    International Nuclear Information System (INIS)

    Schneider, Horst

    2010-01-01

    In Sweden, high-level radioactive waste producing heat, i.e. spent fuel, is to be emplaced for final storage on the site of Forsmark, which also holds three reactor units. The siting decision was taken in June 2009. A 100 percent private company, a merger of the commercial nuclear power plant operators as producers of the waste, is responsible for the siting decision as well as for waste storage. Major impulses were given to the back-end fuel cycle policy in the early 1970s. Sweden practically gave up the reprocessing option very soon, but kept on pursuing final storage in deep geologic formations. Between 1977, when legislation was adopted with conditions relating to repository storage, and 2009, when the decision in favour of the Forsmark site was taken, the path followed was not always a straight line. The boundary conditions, such as the organization of the repository and procedural and safety criteria established by the government, are interesting with regard to their influence on the siting decision, if any. For this reason, the approaches chosen and their connections with government criteria and with geological conditions in Sweden, including their impacts on the repository concept chosen, will be examined. After a summary review of developments in Sweden, filing of the licensing application and the accompanying documents up to commissioning of the repository, a short comparison will be made with the situation in Germany, especially the status reached of the Gorleben salt dome, highlighting and evaluating important criteria and parameters. Sweden as a model is important especially in these respects: A repository site was found by a private company in consensus with the local government within the framework of government criteria, and with ultimate responsibility resting with the government; the local government of a place not winning the siting decision is disappointed although it will have the conditioning plant and receive higher grants; it was not only

  6. Elementary migration around the Oklo nuclear reactors. Implications for high level radioactive wastes storage

    International Nuclear Information System (INIS)

    Menet-Dressayre, C.; Menager, M.T.

    1993-01-01

    The study of Uranium and rare earths near the reactors has displayed the radioelements transfer in the reactors neighbourhood. The main implications for high level radioactive wastes disposal in geological formations are discussed. 12 refs

  7. Handling of spent nuclear fuel and final storage of vitrified high level reprocessing waste

    International Nuclear Information System (INIS)

    1978-01-01

    The report gives a general summary of the Swedish KBS-project on management and disposal of vitrified reprocessed waste. Its final aim is to demostrate that the means of processing and managing power reactor waste in an absolutely safe way, as stipulated in the Swedish so called Conditions Act, already exist. Chapters on Storage facility for spent fuel, Intermidiate storage of reprocessed waste, Geology, Final repository, Transportation, Protection, and Siting. (L.E.)

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

    International Nuclear Information System (INIS)

    Calmus, R.B.

    1998-01-01

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

  9. Lead-iron phosphate glass: a stable storage medium for high-level nuclear waste

    International Nuclear Information System (INIS)

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

    1984-01-01

    Results are presented which show that lead-iron phosphate glasses are a promising new waste form for the safe immobilization of both high-level defense and high-level commercial radioactive waste. Relative to the borosilicate nuclear waste glasses that are currently the ''reference'' waste form for the long-term disposal of nuclear waste, lead-iron phosphate glasses have several distinct advantages: (1) an aqueous corrosion rate that is about 1000 times lower, (2) a processing temperature that is 100 0 to 250 0 C lower and, (3) a much lower melt viscosity in the temperature range from 800 0 to 1000 0 C. Most significantly, the lead-iron phosphate waste form can be processed using a technology similar to that developed for borosilicate nuclear waste glasses

  10. 40 Years of Experience of NIRAS / Belgoprocess on the Interim Storage of Low, Intermediate and High Level Waste

    International Nuclear Information System (INIS)

    Braeckeveldt, Marnix; Ghys, Bart

    2016-01-01

    Conclusion: • ONDRAF/NIRAS and Belgoprocess have gained over time an extended experience on the interim storage of Low-Intermediate and High level waste. • An systematic inspection strategy was developed in order the verify the conformity of the different waste-packages and corrective measures were taken to guarantee safe storage conditions. • From 2022 , ONDRAF/NIRAS will operate a surface disposal facility for LLW

  11. High Level Radioactive Waste Management

    International Nuclear Information System (INIS)

    1991-01-01

    The proceedings of the second annual international conference on High Level Radioactive Waste Management, held on April 28--May 3, 1991, Las Vegas, Nevada, provides information on the current technical issue related to international high level radioactive waste management activities and how they relate to society as a whole. Besides discussing such technical topics as the best form of the waste, the integrity of storage containers, design and construction of a repository, the broader social aspects of these issues are explored in papers on such subjects as conformance to regulations, transportation safety, and public education. By providing this wider perspective of high level radioactive waste management, it becomes apparent that the various disciplines involved in this field are interrelated and that they should work to integrate their waste management activities. Individual records are processed separately for the data bases

  12. Handling of spent nuclear fuel and final storage of vitrified high level reprocessing waste

    International Nuclear Information System (INIS)

    1978-01-01

    A summary of the planning of transportation and plant design in the Swedish KBS project on management and disposal reprocessed radioactive waste. It describes a transportation system, a central storage facility for used fuel elements, a plant for intermediate storage and encapsulation and a final repository for the vitrified waste. Accounts are given for the reprocessing and vitrification. The safety of the entire system is discussed

  13. Handling and storage of high-level liquid wastes from reprocessing of spent fuel

    International Nuclear Information System (INIS)

    Finsterwalder, L.

    1982-01-01

    The high level liquid wastes arise from the reprocessing of irradiated nuclear fuels, which are dissolved in aqueous acid solution, and the plutonium and unburned uranium removed in the chemical separation plant. The remaining solution, containing more than 99% of the dissolved fission products, together with impurities from cladding materials, corrosion products, traces of unseparated plutonium and uranium and most of the transuranic elements, constitutes the high-level waste. At present, these liquid wastes are usually concentrated by evaporation and stored as an aqueous nitric acid solution in high-integrity stainless-steel tanks. There is now world-wide agreement that, for the long term, these liquid wastes should be converted to solid form and much work is in progress to develop techniques for the solidification of these wastes. This paper considers the design requirements for such facilities and the experience gained during nearly 30 years of operation. (orig./RW)

  14. Geochemistry research planning for the underground storage of high-level nuclear waste

    International Nuclear Information System (INIS)

    Apps, J.A.

    1983-09-01

    This report is a preliminary attempt to plan a comprehensive program of geochemistry research aimed at resolving problems connected with the underground storage of high-level nuclear waste. The problems and research needs were identified in a companion report to this one. The research needs were taken as a point of departure and developed into a series of proposed projects with estimated manpowers and durations. The scope of the proposed research is based on consideration of an underground repository as a multiple barrier system. However, the program logic and organization reflect conventional strategies for resolving technological problems. The projects were scheduled and the duration of the program, critical path projects and distribution of manpower determined for both full and minimal programs. The proposed research was then compared with ongoing research within DOE, NRC and elsewhere to identify omissions in current research. Various options were considered for altering the scope of the program, and hence its cost and effectiveness. Finally, recommendations were made for dealing with omissions and uncertainties arising from program implementation. 11 references, 6 figures, 4 tables

  15. Assuring safe interim storage of Hanford high-level tank wastes

    International Nuclear Information System (INIS)

    Bacon, R.F.; Babad, H.; Lerch, R.E.

    1996-01-01

    The federal government established the Hanford Site in South-Eastern Washington near the City of Richland in 1943 to produce plutonium for national defense purposes. The Hanford Site occupies approximately 1,450 square kilometers (560 square miles) of land North of the City of Richland. The production mission ended in 1988, transforming the Hanford Site mission to waste management, environmental restoration, and waste disposal. Thus the primary site mission has shifted from production to the management and disposal of radioactive, hazardous, and mixed waste that exist at the Hanford Site. This paper describes the focus and challenges facing the Tank Waste Remediation System (TWRS) Program related to the dual and parallel missions of interim safe storage and disposal of the tank associated waste. These wastes are presently stored in 2.08E+05 liters (55,000) to 4.16E+06 liters (1,100,000) gallon low-carbon steel tanks. There are 149 single- and 28 double-shell radioactive underground storage tanks, as well as approximately 40 inactive miscellaneous underground storage tanks. In addition, the TWRS mission includes the storage and disposal of the inventory of 1,929 cesium and strontium capsules created as part of waste management efforts. Tank waste was a by-product of producing plutonium and other defense related materials. From 1944 through 1990, four (4) different major chemical processing facilities at the Hanford Site processed irradiated (spent) fuel from defense reactors to separate and recover plutonium for weapons production. As new and improved processes were developed over the last 50 years, the processing efficiency improved and the waste compositions sent to the tanks for storage changed both chemically and radiologically. The earliest separation processes (e.g., bismuth phosphate coprecipitation) carried out in T Plant (1944-1956) and B Plant (1945-1952) recovered only plutonium

  16. REGIONAL BINNING FOR CONTINUED STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTES

    Energy Technology Data Exchange (ETDEWEB)

    W. Lee Poe, Jr

    1998-10-01

    In the Continued Storage Analysis Report (CSAR) (Reference 1), DOE decided to analyze the environmental consequences of continuing to store the commercial spent nuclear fuel (SNF) at 72 commercial nuclear power sites and DOE-owned spent nuclear fuel and high-level waste at five Department of Energy sites by region rather than by individual site. This analysis assumes that three commercial facilities pairs--Salem and Hope Creek, Fitzpatrick and Nine-Mile Point, and Dresden and Moms--share common storage due to their proximity to each other. The five regions selected for this analysis are shown on Figure 1. Regions 1, 2, and 3 are the same as those used by the Nuclear Regulatory Commission in their regulatory oversight of commercial power reactors. NRC Region 4 was subdivided into two regions to more appropriately define the two different climates that exist in NRC Region 4. A single hypothetical site in each region was assumed to store all the SNF and HLW in that region. Such a site does not exist and has no geographic location but is a mathematical construct for analytical purposes. To ensure that the calculated results for the regional analyses reflect appropriate inventory, facility and material degradation, and radionuclide transport, the waste inventories, engineered barriers, and environmental conditions for the hypothetical sites were developed from data for each of the existing sites within the given region. Weighting criteria to account for the amount and types of SNF and HLW at each site were used in the development of the environmental data for the regional site, such that the results of the analyses for the hypothetical site were representative of the sum of the results of each actual site if they had been modeled independently. This report defines the actual site data used in development of this hypothetical site, shows how the individual site data was weighted to develop the regional site, and provides the weighted data used in the CSAR analysis. It is

  17. REGIONAL BINNING FOR CONTINUED STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTES

    International Nuclear Information System (INIS)

    W. Lee Poe, Jr.

    1998-01-01

    In the Continued Storage Analysis Report (CSAR) (Reference 1), DOE decided to analyze the environmental consequences of continuing to store the commercial spent nuclear fuel (SNF) at 72 commercial nuclear power sites and DOE-owned spent nuclear fuel and high-level waste at five Department of Energy sites by region rather than by individual site. This analysis assumes that three commercial facilities pairs--Salem and Hope Creek, Fitzpatrick and Nine-Mile Point, and Dresden and Moms--share common storage due to their proximity to each other. The five regions selected for this analysis are shown on Figure 1. Regions 1, 2, and 3 are the same as those used by the Nuclear Regulatory Commission in their regulatory oversight of commercial power reactors. NRC Region 4 was subdivided into two regions to more appropriately define the two different climates that exist in NRC Region 4. A single hypothetical site in each region was assumed to store all the SNF and HLW in that region. Such a site does not exist and has no geographic location but is a mathematical construct for analytical purposes. To ensure that the calculated results for the regional analyses reflect appropriate inventory, facility and material degradation, and radionuclide transport, the waste inventories, engineered barriers, and environmental conditions for the hypothetical sites were developed from data for each of the existing sites within the given region. Weighting criteria to account for the amount and types of SNF and HLW at each site were used in the development of the environmental data for the regional site, such that the results of the analyses for the hypothetical site were representative of the sum of the results of each actual site if they had been modeled independently. This report defines the actual site data used in development of this hypothetical site, shows how the individual site data was weighted to develop the regional site, and provides the weighted data used in the CSAR analysis. It is

  18. The Centralized Temporary Storage (ATC) Spanish. General and operational readiness in the management of high level radioactive waste

    International Nuclear Information System (INIS)

    Reyes, S.; Diaz, A.; Dilla, A.; Gonzalez, J. R.; Haro, R.

    2014-01-01

    The high-level waste will go to Centralized Temporary Storage (ATC) in Villar de Canas (Cuenca), which is the solution to your management until finished designing its final disposal in the Deep Geological Storage (AGP). Transport containers of high activity radioactive waste will be stored on arrival at the ATC. In time of management, waste will be taken remotely, will be encapsulated in secure conditions and stored in vaults designed to be cooled by natural convection of air confinement. The buildings in the facility will have a general provision optimize the movement of packages and people for the development of all these activities. (Author)

  19. Surface storage of vitrified high-level radioactive waste in reinforced-concrete casks

    International Nuclear Information System (INIS)

    Beale, H.; George, M.W.; Robertson, T.J.M.

    1982-06-01

    The feasibility of storing canisters containing vitrified high level radioactive waste in reinforced concrete casks is examined. This preliminary study identifies the limitations and probable cost of such a store and leads to the conclusion that the concept is feasible. (author)

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

    International Nuclear Information System (INIS)

    1994-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

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

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

    International Nuclear Information System (INIS)

    1994-04-01

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

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

    International Nuclear Information System (INIS)

    1994-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

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

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

    International Nuclear Information System (INIS)

    1994-04-01

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

  7. Social stakes of the reversibility in the deep storage of high level radioactive wastes

    International Nuclear Information System (INIS)

    Heriard-Dubreuil, G.; Schieber, C.; Schneider, T.

    1998-06-01

    This document proposes a study of the conditions which surrounded the reversibility introduction in high activity wastes deep storage at an international scale, as well as a reflexion on the social stakes associated there. In France, the law of december 30, 1991 concerning the research on the radioactive wastes prescribes '' the study of possibilities retrieval or non retrieval storage in deep geological deposits''. The analysis of the reversibility associated social stakes emphasizes the necessity to prevent irreversible consequences, to take care to the choices reversibility, to preserve the future generations autonomy. Thus to elaborate a more satisfactory solution between deep disposal and surface storage, a deep storage, capable of gradually evolution, concept is defined. (A.L.B.)

  8. Method for the conditioning of high level radioactive wastes for their safe storage and disposal

    International Nuclear Information System (INIS)

    Geel, J. van; Eschrich, H.; Detilleux, E.

    1976-01-01

    A method is described for the treatment of solidified high level radioactive wastes to enable them to be safely stored or disposed of in an approved manner. The solidified waste is embedded in a matrix of pure metals or metal alloys. The metals may be Pb, Pb/Sb alloys, Pb/Sn alloys, Pb/Bi alloys, Pb/Zn alloys, or mixtures of these, or Al, Al/Si alloys, Al/Mg alloys, Al/Cu alloys, or mixtures. The matrix is clad with non-corrosive material, selected from stainless steel, Ti, Pb, Pb alloys, Al, Al alloys, or mixtures of same. A non-corrosive container is filled with the solidified waste and is heated to above the melting temperature of the metallic matrix material used to embed the waste. The matrix material is then added and the container is cooled. The container may then be degassed. The solidified waste feed may be in the form of a vitreous material containing the high level waste; this vitreous material may consist of a lead borosilicate or a mixture of non-lead borosilicates and phosphate glasses, and the method of preparing it is described. (U.K.)

  9. Trial storage of high-level waste in the Asse II salt mine

    International Nuclear Information System (INIS)

    1984-01-01

    This report covers a second phase of the work performed by GSF and KfK in the Asse II salt mine, with a view to disposal of radioactive waste in salt formations. New items of the research were geophysical investigations of the behaviour of heated salt and preparation of a trial storage in the Asse II salt mine

  10. High-level waste canister storage final design, installation, and testing. Topical report

    International Nuclear Information System (INIS)

    Connors, B.J.; Meigs, R.A.; Pezzimenti, D.M.; Vlad, P.M.

    1998-04-01

    This report is a description of the West Valley Demonstration Project's radioactive waste storage facility, the Chemical Process Cell (CPC). This facility is currently being used to temporarily store vitrified waste in stainless steel canisters. These canisters are stacked two-high in a seismically designed rack system within the cell. Approximately 300 canisters will be produced during the Project's vitrification campaign which began in June 1996. Following the completion of waste vitrification and solidification, these canisters will be transferred via rail or truck to a federal repository (when available) for permanent storage. All operations in the CPC are conducted remotely using various handling systems and equipment. Areas adjacent to or surrounding the cell provide capabilities for viewing, ventilation, and equipment/component access

  11. High-level waste canister storage final design, installation, and testing. Topical report

    Energy Technology Data Exchange (ETDEWEB)

    Connors, B.J.; Meigs, R.A.; Pezzimenti, D.M.; Vlad, P.M.

    1998-04-01

    This report is a description of the West Valley Demonstration Project`s radioactive waste storage facility, the Chemical Process Cell (CPC). This facility is currently being used to temporarily store vitrified waste in stainless steel canisters. These canisters are stacked two-high in a seismically designed rack system within the cell. Approximately 300 canisters will be produced during the Project`s vitrification campaign which began in June 1996. Following the completion of waste vitrification and solidification, these canisters will be transferred via rail or truck to a federal repository (when available) for permanent storage. All operations in the CPC are conducted remotely using various handling systems and equipment. Areas adjacent to or surrounding the cell provide capabilities for viewing, ventilation, and equipment/component access.

  12. Handling of spent nuclear fuel and final storage of nitrified high level reprocessing waste

    International Nuclear Information System (INIS)

    The following stages of handling and transport of the fuel on its way to final storage are dealt with in the report. 1) The spent nuclear fuel is stored at the power station or in the central fuel storage facility awaiting reprocessing. 2) The fuel is reprocessed, i.e. uranium, plutonium and waste are separated from each other. Reprocessing does not take place in Sweden. The highlevel waste is vitrified and can be sent back to Sweden in the 1990s. 3) Vitrified waste is stored for about 30 years awaiting deposition in the final repository. 4) The waste is encapsulated in highly durable materials to prevent groundwater from coming into contact with the waste glass while the radioactivity of the waste is still high. 5) The canisters are emplaced in a final repository which is built at a depth of 500 m in rock of low permeability. 6) All tunnels and shafts are filled with a mixture of clay and sand of low permeability. A detailed analysis of possible harmful effects resulting from normal acitivties and from conceivable accidents is presented in a special section. (author)

  13. The chemical durability of glasses suitable for the storage of high level radioactive wastes, (1)

    International Nuclear Information System (INIS)

    Terai, Ryohei; Hara, Shigeo; Kawamoto, Takamichi; Nanbu, Tadahiko; Nakamura, Takao.

    1975-01-01

    To develop the glassy materials suitable for the long-term storage of high level radioactive wastes, the chemical durability of the glasses of borax-alumina-silica system has been investigated. The test was carried out by the following three ways, (1) glass-disk immersion method, (2) continuous leach method and (3) method prescribed in JIS-R3502. In the continuous leach method, glass grains were exposed to circulating water at a constant temperature for a week to obtain the leach factor or leach rate. It was found from the experimental results that, as the silica content increased, the melting temperature of the glasses progressively increased and the chemical durability was considerably improved, and that B 2 O 3 and Na 2 O constituents were preferentially dissolved in water leaving relatively insoluble components such as SiO 2 and Al 2 O 3 . The rate at which B 2 O 3 and Na 2 O in glass are leached out is governed by three processes, that is, (1) the boundary reaction on the glass surface, (2) the diffusion process through the hydrated layer, and (3) the disintegration of hydrated layer. The first process probably corresponds to the hydration of boric oxides on the glass surface or to the ion exchange between protons in solution and Na + ions in glass, and the second process seems to correspond to the diffusion of protons through the hydrated layer on the glass surface. Although the ratio of [Na-BO 4 ]/[BO 3 ] in the borax-silica glasses was determined to be 0.5 by means of NMR measurement, Na 2 O/B 2 O 3 ratio in leached solution was less than 0.5, indicating that [BO 3 ] groups in glass were more soluble than [Na-BO 4 ] groups. From the viewpoint of appreciation of safety, the chemical durability of the glasses of borax-aluminasilica system was rather unsatisfactory, but that of the glasses containing silica in quantities was comparable to the soda-lime silicate sheet glasses. (auth.)

  14. Multilayer Protective Coatings for High-Level Nuclear Waste Storage Containers

    Science.gov (United States)

    Fusco, Michael

    Corrosion-based failures of high-level nuclear waste (HLW) storage containers are potentially hazardous due to a possible release of radionuclides through cracks in the canister due to corrosion, especially for above-ground storage (i.e. dry casks). Protective coatings have been proposed to combat these premature failures, which include stress-corrosion cracking and hydrogen-diffusion cracking, among others. The coatings are to be deposited in multiple thin layers as thin films on the outer surface of the stainless steel waste basket canister. Coating materials include: TiN, ZrO2, TiO2, Al 2O3, and MoS2, which together may provide increased resistances to corrosion and mechanical wear, as well as act as a barrier to hydrogen diffusion. The focus of this research is on the corrosion resistance and characterization of single layer coatings to determine the possible benefit from the use of the proposed coating materials. Experimental methods involve electrochemical polarization, both DC and AC techniques, and corrosion in circulating salt brines of varying pH. DC polarization allows for estimation of corrosion rates, passivation behavior, and a qualitative survey of localized corrosion, whereas AC electrochemistry has the benefit of revealing information about kinetics and interfacial reactions that is not obtainable using DC techniques. Circulation in salt brines for nearly 150 days revealed sustained adhesion of the coatings and minimal weight change of the steel samples. One-inch diameter steel coupons composed of stainless steel types 304 and 316 and A36 low alloy carbon steel were coated with single layers using magnetron sputtering with compound targets in an inert argon atmosphere. This resulted in very thin films for the metal-oxides based on low sputter rates. DC polarization showed that corrosion rates were very similar between bare and coated stainless steel samples, whereas a statistically significant decrease in uniform corrosion was measured on coated

  15. Bentonite-like material sealing to high-level radioactive wastes storage

    International Nuclear Information System (INIS)

    Linares, J.; Linares Gonzalez, J.; Huertas Garcia, F.; Reyes Camacho.

    1993-01-01

    Among the most used materials for sealing of radioactive waste storage, bentonite shows a high number of advantages because of its plasticity, thermal and hydraulic conductivity, etc. The paper makes a review on different Spanish deposits of bentonite and their stability. Most of studies are focussed on the volcanic region at Cabo de Gata (Almeria). That area offers the most productive hydrothermal bentonite deposits in Spain

  16. Heat transfer in high-level waste management

    International Nuclear Information System (INIS)

    Dickey, B.R.; Hogg, G.W.

    1979-01-01

    Heat transfer in the storage of high-level liquid wastes, calcining of radioactive wastes, and storage of solidified wastes are discussed. Processing and storage experience at the Idaho Chemical Processing Plant are summarized for defense high-level wastes; heat transfer in power reactor high-level waste processing and storage is also discussed

  17. Design and operation of high level waste vitrification and storage facilities

    International Nuclear Information System (INIS)

    1992-01-01

    The conversion of high level wastes (HLW) into solids has been studied for the past 30 years, primarily in those countries engaged in the reprocessing of nuclear fuels. Production and demonstration calcination and solidification plants have been operated by using waste solutions from fuels irradiated at various burnup rates, depending on the reactor type. Construction of more advanced solidification processes is now in progress in several countries to permit the handling of high burnup power reactor fuel wastes. The object of this report is to provide detailed information and references for those vitrification systems in advanced stages of implementation. Some less detailed information will be provided for previously developed immobilization systems. The report will examine the HLLW arising from the various locations, the features of each process as well as the stage of development, scale-up potential and flexibility of the processes. Since the publication of IAEA Technical Reports Series No. 176, Techniques for the Solidification of High-Level Wastes great progress on this subject has been made. The AVM in France has been operated successfully for 11 years and France has completed construction at La Hague of two vitrification plants that are based on the AVM rotary calciner/metallic melter process. A similar plant is under construction at Sellafield. The ceramic melter process has been chosen by several countries. Germany has successfully operated the PAMELA vitrification plant. Since 1986, Belgoprocess has continued to operate this facility. The former USSR operated the EP-500 plant from 1986 to 1988. In addition, two ceramic melter vitrification plants are nearing completion in the USA at Savannah River and West Valley and plans are being made to use this technology at Hanford as well as in Japan, Germany and India. This major progress attests to the maturity of these technologies for vitrifying HLLW to make a borosilicate glass for disposal of the waste. 67

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

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

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

    International Nuclear Information System (INIS)

    1994-04-01

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

  20. Studies of corrosion in metallic container for storage of high level radioactive wastes

    International Nuclear Information System (INIS)

    Azkarate, I.; Madina, V.; Insausti, M.

    1999-01-01

    The metallic container is one of the most important barriers that, along with engineered and natural barriers, will isolate high level nuclear waste in saline and granite geological formations from the geosphere. However, general and localized corrosion modes such as stress corrosion cracking (SCC), pitting, crevice corrosion and hydrogen damage can be active under disposal conditions, so the corrosion behaviour of the metal container material must be carefully studied. Several metals and their alloys have been proposed for the fabrication of nuclear waste containers including carbon steels, stainless steels, titanium and titanium alloys and copper and copper-base alloys. Carbon steels and copper alloys are considered for the two rock formations, titanium is considered for salt environments and the stainless steel only in the case of a granite formation. (Author)

  1. Assessment of degradation concerns for spent fuel, high-level wastes, and transuranic wastes in monitored retrievalbe storage

    International Nuclear Information System (INIS)

    Guenther, R.J.; Gilbert, E.R.; Slate, S.C.; Partain, W.L.; Divine, J.R.; Kreid, D.K.

    1984-01-01

    It has been concluded that there are no significant degradation mechanisms that could prevent the design, construction, and safe operation of monitored retrievable storage (MRS) facilities. However, there are some long-term degradation mechanisms that could affect the ability to maintain or readily retrieve spent fuel (SF), high-level wastes (HLW), and transuranic wastes (TRUW) several decades after emplacement. Although catastrophic failures are not anticipated, long-term degradation mechanisms have been identified that could, under certain conditions, cause failure of the SF cladding and/or failure of TRUW storage containers. Stress rupture limits for Zircaloy-clad SF in MRS range from 300 to 440 0 C, based on limited data. Additional tests on irradiated Zircaloy (3- to 5-year duration) are needed to narrow this uncertainty. Cladding defect sizes could increase in air as a result of fuel density decreases due to oxidation. Oxidation tests (3- to 5-year duration) on SF are also needed to verify oxidation rates in air and to determine temperatures below which monitoring of an inert cover gas would not be required. Few, if any, changes in the physical state of HLW glass or canisters or their performance would occur under projected MRS conditions. The major uncertainty for HLW is in the heat transfer through cracked glass and glass devitrification above 500 0 C. Additional study of TRUW is required. Some fraction of present TRUW containers would probably fail within the first 100 years of MRS, and some TRUW would be highly degraded upon retrieval, even in unfailed containers. One possible solution is the design of a 100-year container. 93 references, 28 figures, 17 tables

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

    International Nuclear Information System (INIS)

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

    1991-09-01

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

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

    International Nuclear Information System (INIS)

    Conrads, T.J.

    1993-06-01

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

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

    International Nuclear Information System (INIS)

    Conrads, T.J.

    1993-01-01

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

  5. Prediction of temperature increases in a salt repository expected from the storage of spent fuel or high-level waste

    International Nuclear Information System (INIS)

    Llewellyn, G.H.

    1978-04-01

    Comparisons in temperature increases incurred from hypothetical storage of 133 MW of 10-year-old spent fuel (SF) or high-level waste (HLW) in underground salt formations have been made using the HEATING5 computer code. The comparisons are based on far-field homogenized models that cover areas of 65 and 25 sq miles for SF and HLW, respectively, and near-field unit-cell models covering respective areas of 610 ft 2 and 400 ft 2 . Preliminary comparisons based on heat loads of 150 kW/acre and 3.5 kW/canister indicated near-field temperature increases about 20% higher for the storage of the spent fuel than for the high-level waste. In these comparisons, it was also found that the thermal energy deposited in the salt after 500 years is about twice the energy deposited by the high-level waste. The thermal load in a repository containing 10-year-old spent fuel was thus limited to 60 kW/acre to obtain comparable far-field thermal effects as obtained in a repository containing 10-year-old high-level waste loaded at 150 kW/acre. Detailed far-field and unit-cell comparisons of transient temperature increases have been made based on these loadings. Unit-cell comparisons were made between a canister containing high-level waste with an initial heat production rate of 2.1 kW and a canister containing a PWR spent fuel assembly producing 0.55 kW. Using a three-dimensional unit-cell model, a maximum salt temperature increase of 260 0 F was calculated for the high-level waste prior to back-filling (5 years after burial), whereas a maximum temperature increase of 110 0 F was calculated for the spent fuel prior to backfilling (25 years after burial). Comparisons were also made between various configurational models for the high-level waste showing the applicability of each model

  6. Update on ASME rules for spent nuclear fuel and high level radioactive material and waste storage containments

    International Nuclear Information System (INIS)

    Ralph S. Hill III; Foster, G.M.

    2005-01-01

    In 2004, a new Code Case, N-717, of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) was published. The Code Case provides rules for construction of containments used for storage of spent nuclear fuel and high level radioactive material and waste. The Code Case has been incorporated into Section III of the Code as Division 3, Subsection WC, Class SC Storage Containments, and will be published in the 2005 Addenda. This paper provides an informative background and insight for these rules to provide Owners, regulators, designers, and fabricators with a more comprehensive understanding of the technical basis for these rules. (authors)

  7. Review of Analytes of Concern and Sample Methods for Closure of DOE High Level Waste Storage Tanks

    International Nuclear Information System (INIS)

    Thomas, T.R.

    2002-01-01

    Sampling residual waste after tank cleaning and analysis for analytes of concern to support closure and cleaning targets of large underground tanks used for storage of legacy high level radioactive waste (HLW) at Department of Energy (DOE) sites has been underway since about 1995. The DOE Tanks Focus Area (TFA) has been working with DOE tank sites to develop new sampling plans, and sampling methods for assessment of residual waste inventories. This paper discusses regulatory analytes of concern, sampling plans, and sampling methods that support closure and cleaning target activities for large storage tanks at the Hanford Site, the Savannah River Site (SRS), the Idaho National Engineering and Environmental Laboratory (INEEL), and the West Valley Demonstration Project (WVDP)

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

    International Nuclear Information System (INIS)

    Allen, G.K.

    1980-08-01

    Near-field temperatures resulting from the storage of high-level waste canisters and spent unreprocessed fuel assembly canisters in geologic formations were determined. Preliminary design of the repository was modeled for a heat transfer computer code, HEATING5, which used the Crank-Nicolson finite difference method to evaluate transient heat transfer. The heat transfer system was evaluated with several two- and three-dimensional models which transfer heat by a combination of conduction, natural convention, and radiation. Physical properties of the materials in the model were based upon experimental values for the various geologic formations. The effects of canister spacing, fuel age, and use of an overpack were studied for the analysis of the spent fuel canisters; salt, granite, and basalt were considered as the storage media for spent fuel canisters. The effects of canister diameter and use of an overpack were studied for the analysis of the high-level waste canisters; salt was considered as the only storage media for high-level waste canisters. Results of the studies on spent fuel assembly canisters showed that the canisters could be stored in salt formations with a maximum heat loading of 134 kw/acre without exceeding the temperature limits set for salt stability. The use of an overpack had little effect on the peak canister temperatures. When the total heat load per acre decreased, the peak temperatures reached in the geologic formations decreased; however, the time to reach the peak temperatures increased. Results of the studies on high-level waste canisters showed that an increased canister diameter will increase the canister interior temperatures considerably; at a constant areal heat loading, a 381 mm diameter canister reached almost a 50 0 C higher temperature than a 305 mm diameter canister. An overpacked canister caused almost a 30 0 C temperature rise in either case

  9. Experimental storage of high-level radioactive wastes in the Asse salt mine - technical aspects

    International Nuclear Information System (INIS)

    Gies, H.; Rothfuchs, T.; Feddersen, H.; Graefe, V.; Gross, S.; Hente, B.; Jockwer, N.; Kessels, W.; Schwaegermann, H.

    1988-01-01

    The work performed under this project in the Asse salt mine is an important milepost within the framework schedule of the 'Gorleben Poject' of Physikalisch-Technische Bundesanstalt (PTB). The project phase I (1982 - June 30, 1985) is about to be concluded at the time this report is published. The main points of interest of this project phase cover the planning of the experimental work, the design of experiments, and the first activities for developing the systems for handling the high-level radioactive wastes. The engineering development work has been advanced to the point where construction and manufacture of equipment can be started (transport containers Asse, TB1, collective transport containers, borehole gates, transport vehicles, waste positioning equipment, and borehole casing). Testing of the pipes for the last mentioned task with regard to the material's deformation behaviour will be done by the Dutch ECN as a sub-contractor. First laboratory experiments have been carried out on radiolysis gas formation, to complement the engineering work and the in-situ measuring programmes. (orig./RB) [de

  10. Numerical modelling of the long term mechanical behaviour of a storage cell for high level nuclear wastes

    International Nuclear Information System (INIS)

    Bian, H.B.; Jia, Y.; Duveau, G.; Shao, J.F.; Poutrel, A.

    2010-01-01

    Document available in extended abstract form only. In France, a clay formation, called Callovo-Oxfordian clay, is considered as potential geological barrier for storage of radioactive wastes. An underground research laboratory is constructed in this clay layer in Bure (CMHM). The French Radioactive Waste Management Agency (Andra) plans to conduct a series of full scale experiments in the Bure Underground Research Laboratory in order to study the coupled Chemo- Thermo-Hydraulic-Mechanical processes due to the excavation, ventilation, subsequent backfilling/sealing, corrosion of metal components and heat release from the radioactive wasters. These phenomena will occur in the storage structures and the geologic barrier for a short/very long period of time. Especially for the High-Level Wastes, due to the important increase of temperature released by the waste packages, the behaviour of geological barrier is perhaps perturbed and the design of repository is also influenced in terms of admissible thermal loading. In this paper, a disposal cell for the radioactive High-Level Wastes (HLW) is studied in order to analyser the coupled phenomena evolving in the HLW storage. The carbon steel is adopted as the first choice material for the container of waste because of its relevant geochemical condition. In the underground radioactive wastes storage, due to the arrival of water from the geological formation, the metallic components will be corroded, by step, with a decrease of mechanical resistance and a generated supplementary volume. The space between the metallic container and the host rock will then be closed under the combination of metallic corrosion and the behaviour differs of Callovo-Oxfordian clay. As results, important stress will be created on the interface of Callovo-Oxfordian clay and wastes packages. In addition, local hydration of barrier modifies the water infiltration from the Callovo-Oxfordian clay and the stress distribution in the argillite. According to

  11. Geochemistry of actinides. Application to the storage of high level radioactive wastes. Under the supervision of Mr Michel Treuil

    International Nuclear Information System (INIS)

    Bouabdallah, Noureddine; Cunault, Jean-Baptiste; Houtin, Gwenaelle; Leborgne, Francois; Lemaire, Celine; Lemarchand, Damien; Quitte, Ghylaine

    1997-06-01

    This collective research report first addresses the chemistry of actinides with a description of their atomic orbitals and the study of their behaviour in solution. The author addresses several aspects: historical overview on actinides, radioactivity, chemical reactions in aqueous solution, redox chemistry, speciation in solution with respect to water characteristics in deep storage conditions. The second part gathers several studies performed on a natural laboratory (the Oklo site in which nuclear reactions occurred about 2 billions years ago) and reports the modelling of radionuclide transfer within a geological system (the model is applied to the Oklo site). The third part addresses issues related to the nuclear fuel cycle, and the storage modes and materials envisaged and involved regarding the storage of high level radioactive wastes, notably in France

  12. Transport of a solute pulse through the bentonite barrier of deep geological high-level waste storage facilities in granite

    International Nuclear Information System (INIS)

    Cormenzana Lopez, J.L.; Alonso Diaz-Teran, J.; Gonzalez- Herranz, E.

    1997-01-01

    Spain like Sweden, Finland, Canada and other countries has opted for an open nuclear fuel cycle, and to store the unreprocessed spent fuel in a stable geological formation. Sweden, Finland and Canada have chosen granite rock for their high-level waste storage facilities. Their Performance Assessment of disposal systems have all obtained to the same result. The greatest annual doses are caused by I 129 in the gap between the fuel rods and the cladding. The reference concept for the Spanish high-level waste storage facility in granite provides for final storage in a granite mass at a depth of 500 m in carbon steel capsules in horizontal tunnels surrounded by a bentonite buffer. It the capsule fails due to generalised corrosion, an not giving credit for the cladding, the I 129 and other radionuclides in the gap would pass immediately into the surrounding water. This paper describes the modelling of the transport of the solute through the bentonite around the capsule to determine the fraction that crosses the bentonite each year. It also analyses the sensitivity of the results to the boundary condition adopted and changes in the values of the relevant parameters. (Author)

  13. Action plan for response to abnormal conditions in Hanford high level radioactive liquid waste storage tanks containing flammable gases

    International Nuclear Information System (INIS)

    Sherwood, D.J.

    1994-03-01

    Radioactive liquid waste tends to produce hydrogen as a result of the interaction of gamma radiation and water. In tanks containing organic chelating agents, additional hydrogen gas as well as nitrous oxide and ammonia can be produced by thermal and radiolytic decomposition of these organics. Several high-level radioactive liquid waste storage tanks, located underground at the Hanford Site, contain waste that retains the gases produced in them until large quantities are released rapidly to the tank vapor space. Tanks filled to near capacity have relatively little vapor space; therefore, if the waste suddenly releases a large amount of hydrogen and nitrous oxide, a flammable gas mixture may result. The most notable waste tank with a flammable gas problem is tank 241-SY-101. Waste in this tank has occasionally released enough flammable gas to burn if an ignition source had been present inside of the tank. Several other waste tanks exhibit similar behavior to a lesser magnitude. Administrative controls have been developed to assure that these Flammable Gas Watch List tanks are safely maintained. Responses have also been developed for off-normal conditions which might develop in these tanks. In addition, scientific and engineering studies are underway to further understand and mitigate the behavior of the Flammable Gas Watch List tanks

  14. Analysis of factors influencing the reliability of retrievable storage canisters for containment of solid high-level radioactive waste

    International Nuclear Information System (INIS)

    Mecham, W.J.; Seefeldt, W.B.; Steindler, M.J.

    1976-08-01

    The reliability of stainless steel type 304L canisters for the containment of solidified high-level radioactive wastes in the glass and calcine forms was studied. A reference system, drawn largely from information furnished by Battelle Northwest Laboratories and Atlantic Richfield Hanford Company is described. Operations include filling the canister with the appropriate waste form, interim storage at a reprocessing plant, shipment in water to a Retrievable Surface Storage Facility (RSSF), interim storage at the RSSF, and shipment to a final disposal facility. The properties of stainless steel type 304L, fission product oxides, calcine, and glass were reviewed, and mechanisms of corrosion were identified and studied. The modes of corrosion important for reliability were stress-corrosion cracking, internal pressurization of the canister by residual impurities present, intergranular attack at the waste-canister interface, and potential local effects due to migration of fission products. The key role of temperature control throughout canister lifetime is considered together with interactive effects. Methods of ameliorating adverse effects and ensuring high reliability are identified and described. Conclusions and recommendations are presented

  15. High-level-waste immobilization

    International Nuclear Information System (INIS)

    Crandall, J.L.

    1982-01-01

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

  16. Combined storage system for LWR spent fuel and high-level waste

    International Nuclear Information System (INIS)

    Baxter, B.J.; Ganley, J.T.; Washington, J.A.

    1983-01-01

    The MODREX storage system consists of four basic elements: (1) the storage canister, (2) the storage module, (3) the storage cask, and (4) the transport cask. The storage canister is the heart of the system and, when used in combination with the module or either of the casks, allows the MODREX system to respond quickly to varied storage system requirements. The MODREX system can be used to hold either spent fuel assemblies or canistered solidified HLW. The ability to combine a basic storage canister with either a concrete module or a metal cask provides flexibility to meet a wide range of storage requirements. The spent fuel is stored in a dry, inert atmosphere, which essentially eliminates corrosion or deterioration of the cladding during extended storage periods. The storage canister and concrete storage module provide additional barriers against radioactivity release, enhancing long-term safety. Heat dissipation is passive, eliminating the need for additional emergency cooling systems or special redundancy. Modular, expandable construction permits minimum initial investment and capital carrying charges; additional capacity is built and paid for only as it is needed, retaining flexibility. 6 references, 2 figures, 1 table

  17. High-Level Radioactive Waste.

    Science.gov (United States)

    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…

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  19. Environmental permits and approvals plan for high-level waste interim storage, Project W-464

    International Nuclear Information System (INIS)

    Deffenbaugh, M.L.

    1998-01-01

    This report discusses the Permitting Plan regarding NEPA, SEPA, RCRA, and other regulatory standards and alternatives, for planning the environmental permitting of the Canister Storage Building, Project W-464

  20. Dry storage facility for spent fuel or high-level wastes

    International Nuclear Information System (INIS)

    Geoffroy, J.; Dobremelle, M.; Fabre, J.C.; Bonnet, C.

    1989-01-01

    The French Atomic Energy Commission (CEA) has specific irradiated fuels which, due to their properties, cannot be reprocessed directly in existing industrial facilities. Accordingly, for the spent fuels from the EL4 and OSIRIS power plants, the CEA has been faced with the problem of selecting a process that will allow the storage of these materials under satisfactory technical and economic conditions. The authors discuss how three conditions must be satisfied to store irradiated fuels releasing heat: containment of radioactive materials, biological shielding, and thermal cooling to guarantee an acceptable temperature- level throughout. In view of the need for an interim storage facility using a simple cooling process requiring only minimal maintenance and monitoring, dry storage in a concrete vault cooled by natural convection was selected. This choice was made within the framework of a research and development program in which theoretical heat transfer investigations and mock-up tests confirmed the feasibility of cooling by natural convection

  1. The underground retrievable storage (URS) high-level waste management concept

    International Nuclear Information System (INIS)

    Ramspott, L.D.

    1991-01-01

    This papers presents the concept of long-term underground retrievable storage (URS) of spent reactor fuel in unsaturated rock. Emplacement would be incremental and the system is planned to be experimental and flexible. The rationale for retrievability is examined, and a technical basis for 300-year retrievability is presented. Maximum isolation is the rationale for underground as opposed to surface storage. Although the potential repository site at Yucca Mountain Nevada would be suitable for a URS, alternate sites are discussed. The technical issues involved in licensing a URS for 300 years are simpler than licensing a 10,000 year repository. 16 refs

  2. Radiolysis salt phenomenology: application to storage of high level radioactive waste

    International Nuclear Information System (INIS)

    Akram, Najib

    1993-01-01

    In France, rock salt is a candidate repository for highly radioactive waste. Rock salt contains water and adsorbed gases which can be released in boreholes after heating due to vitrified wastes. In addition, waste-induced irradiation in near-field conditions induce radiolytic reactions which also contribute to gas release. The aim of this work is to understand and evaluate the effects of heat and irradiation produced by waste containers in a deep disposal, primarily concerning gas production. This is justified by the impact of gases on long-term safety: toxicity, explosibility, chemical reactivity, pressure build-up. We have evidenced the influence of integrated dose, filling gases, temperature and grain size on an homogeneous medium (Asse Mine rock salt). We have then studied heterogeneous samples, which allowed to determine the influence of the chemical and mineralogical composition of rock salt (bedded rock salt from the Mine de Potasse d'Alsace). The role played by organic matter on gas production is important, leading for instance to high consumption rates of oxygen. Through this study, we have also considered the behaviour of clay-rich materials under irradiation. Our results constitute important bases for the future modelling of the phenomena which will take place in the near-field of a rock salt-type repository, especially concerning its long-term safety. (author) [fr

  3. Studies for the storage of high level and long living radioactive wastes

    International Nuclear Information System (INIS)

    2001-01-01

    This synthesis presents the studies and main results of the year 1999 obtained in the framework of the second axis of the 1991 law about the feasibility study of deep underground radioactive waste repositories. It precises also the French and international context in which researches are carried out, in particular with the development of scientific partnerships. (J.S.)

  4. Radiation exposure rate and liquid level measurement inside a high level liquid waste (HLLW) storage tank

    International Nuclear Information System (INIS)

    Sur, B.; Yue, S.; Thekkevarriam, A.

    2007-01-01

    An instrument based on an inexpensive, small silicon diode has been developed and used to measure, for the first time, the gamma radiation exposure rate profile inside a 6.4 mm diameter reentrant thermo-well tube, immersed in the highly radioactive liquid solution in an HLLW storage tank. The measurement agrees with previous calculations of exposure rate, and provides confirmation for safe and effective radiation work plans and material selection for investigations and remediation of the storage tank facility. The measured radiation exposure rate profile is also used to confirm that the position of tank internal structures have not changed because of aging and corrosion, and to obtain, within a few mm, the level of liquid inside the tank. (author)

  5. Operations and Maintenance Concept Plan for the Immobilized High Level Waste (IHLW) Interim Storage Facility

    Energy Technology Data Exchange (ETDEWEB)

    JANIN, L.F.

    2000-08-30

    This O&M Concept looks at the future operations and maintenance of the IHLW/CSB interim storage facility. It defines the overall strategy, objectives, and functional requirements for the portion of the building to be utilized by Project W-464. The concept supports the tasks of safety basis planning, risk mitigation, alternative analysis, decision making, etc. and will be updated as required to support the evolving design.

  6. Operations and Maintenance Concept Plan for the Immobilized High-Level Waste (IHLW) Interim Storage Facility

    International Nuclear Information System (INIS)

    JANIN, L.F.

    2000-01-01

    This OandM Concept looks at the future operations and maintenance of the IHLW/CSB interim storage facility. It defines the overall strategy, objectives, and functional requirements for the portion of the building to be utilized by Project W-464. The concept supports the tasks of safety basis planning, risk mitigation, alternative analysis, decision making, etc. and will be updated as required to support the evolving design

  7. Radiant energy dissipation during final storage of high-level radioactive waste in rock salt

    International Nuclear Information System (INIS)

    Ramthun, H.

    1981-08-01

    A final disposal concept is assumed where the high-active waste from 1400 t of uranium, remaining after conditioning, is solidified in borosilicate glass and distributed in 1.760 waste casks. These containers 1.2 m in height and 0.3 m in diameter are to be buried 10 years after the fuel is removed from the reactor in the 300 m deep boreholes of a salt dome. For this design the mean absorbed dose rates are calculated in the glass die (3.9 Gy/s), the steel mantle (0.26 Gy/s) and in the salt rock (0.12 Gy/s at a distance of 1 cm and 0.034 Gy/s at a distance of 9 cm from the container surface) valid at the beginning of disposal. The risk involved with these amounts of stored lattice energy is shortly discussed. (orig.) [de

  8. Preliminary design of the high-level waste canister storage system: Topical report for the period of January 1, 1987--September 30, 1987

    International Nuclear Information System (INIS)

    Peters, F.E.; Leap, D.R.

    1987-11-01

    The final stage of the West Valley solidification program will be to place the high-level waste canisters in interim storage until a federal repository is ready to receive them. The waste canisters will be stored in the largest former fuel reprocessing cell at West Valley modified for this purpose. This report provides a description of the preliminary design of the Waste Canister Storage Facility. 9 refs., 14 figs., 1 tab

  9. Thermal stresses in a repository for ultimate storage of high-level radioactive wastes

    International Nuclear Information System (INIS)

    Ehlert, C.

    1981-01-01

    An important factor to be considered in evaluating the suitability of a salt mine as a waste repository is the deformation behaviour of rock salt, as this is the predominant type of rock in this formation. Equations are presented and explained describing the elastic, plastic, and viscoplastic deformation mechanisms contributing to overall rock salt deformation, and use of these equations is made through a specially developed arithmetic method. As there are stratifications and discontinuties in the formation to be considered in the computation, additional criteria are to be taken into account in the integrity considerations, especially the figures of material equations for all other types of rock occurring in the formation. (DG) [de

  10. Systems engineering management and implementation plan for Project W-464, immobilized high-level waste storage

    International Nuclear Information System (INIS)

    Wecks, M.D.

    1998-01-01

    The Systems Engineering Management and Implementation Plan (SEMIP) for TWRS Project W-46 describes the project implementation of the Tank Waste Remediation System Systems Engineering Management Plan. (TWRS SEMP), Rev. 1. The SEMIP outlines systems engineering (SE) products and processes to be used by the project for technical baseline development. A formal graded approach is used to determine the products necessary for requirements, design, and operational baseline completion. SE management processes are defined, and roles and responsibilities for management processes and major technical baseline elements are documented

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

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

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

    International Nuclear Information System (INIS)

    1994-04-01

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

  13. Trial storage of high-level waste cylinders in the Asse II salt mine

    International Nuclear Information System (INIS)

    1984-01-01

    This report covers the contract period 1976-77, as well as some of the tasks carried out during the extension in 1978, in the framework of the R and D programme for disposal of radioactive waste in salt formations. With regard to the in-situ tests for the liberation and migration of brine, the testing devices were examined successfully. Laboratory examinations carried out showed a stepwise liberation of the water contents in halite in dependence on the temperature. The amount of brine liberated stood in good agreement with the in situ results. A temperature test for borehole convergence resulted in definite convergence rates. Simultaneously no influence was registered in the stability of the surrounding rocks. For the realization of an integrated major experiment, temperature test field IV was mined on the 750 m level of the Asse Salt Mine and heater- as well as measurement drillings were carried out. Extensive rheological examinations are concentrated particularly on the halite and secondly on the Carnallite. They are chiefly based on uni- and multiaxial pressure tests. Computer programmes are developed to examine the heat generation in wastes as well as in salt. In comparison, the programme development of computer codes for the stability behaviour of rocks is still at a relatively early stage, because it has to build up on the results of heat generation. The works for the development of a transport container with a shielding combination are at a very advanced stage. An integrated disposal- and retrieval system was developed, tested and successfully demonstrated. A monitoring system in the mine has also been developed in its essential parts

  14. CASTOR registered HAW28M - a high heat load cask for transport and storage of vitrified high level waste containers

    International Nuclear Information System (INIS)

    Vossnacke, A.; Klein, K.; Kuehne, B.

    2004-01-01

    Within the German return programme for vitrified high level waste (HLW) from reprocessing at COGEMA and BNFL up to now 39 casks loaded with 28 containers each were transported back to Germany and are stored in the Interim Storage Facility Gorleben (TBL-G) for up to 40 years. For transport and storage in all but one case the GNB casks CASTOR registered HAW 20/28 CG have been used. This cask type is designed to accommodate 20 or 28 HLW containers with a total thermal power of 45 kW maximum. In the near future, among the high level waste, which has to be returned to Germany, there will be an increasing number of containers of which the heat capacity and radioactive inventory will exceed the technical limits of the CASTOR registered HAW 20/28 CG. Therefore GNB has started the development of a new cask generation, named CASTOR registered HAW28M, meeting these future requirements. The CASTOR registered HAW28M is especially developed for the transport of vitrified residues from France and Great Britain to Germany. It complies with the international regulations for type B packages according to IAEA (International Atomic Energy Agency). It is thus guaranteed that even in case of any accident the cask body and the lid system remain functional and the safe confinement of the radioactive contents remains intact during transport. The CASTOR registered HAW28M fulfills not only the requirements for transport but also the acceptance criteria of interim storage: radiation shielding, heat dissipation, safe confinement under both normal and hypothetical accident conditions. Storage buildings such as the TBL-G simply support the safety functions of the cask. The challenge for the development results from higher requirements of the technical specification, particularly related to fuel which is reprocessed. As a consequence of the reprocessing of fuel with increased enrichment and burn up, higher heat capacity and sophisticated shielding measures have to be considered. For the CASTOR

  15. High level radioactive wastes storage characterization and long-term behaviour of spent fuels

    International Nuclear Information System (INIS)

    Diaz Arocas, P.P.; Garcia Serrano, J.; Mendez Martin, F.J.; Quinones Diez, J.; Rodriguez Almazan, J.L.; Serrano Agejas, J.A.; Esteban Hernandez, J.A.

    1997-04-01

    The knowledge of long term spent fuel behaviour in a repository is one of the main goals in the waste management assessment due to its influence on repository design topics and on the performance assessment. At the moment, Spain has not selected a geological formation for a final repository. Therefore, R AND D activities are performed by considering granite, salt and clay as candidate options. This report summarises the activities carried out in CIEMAT from 1991 to 1995 in the frame of the Agreement between CIEMAT and ENRESA in the Area of spent fuel direct disposed. Experimental activities include leaching experiments of spent fuel, UO 2 and SIMFUEL and co-precipitation/solubility experiments of relevant secondary solid phases expected under repository conditions. The objective of leaching studies is to understand the processes which will occur when the underground water accede to the source term and to provide leaching rates of spent fuel and the influence of several variables as pH, Eh, etc. The co-precipitation/solubility experiments are focused on the knowledge of the formation conditions of relevant secondary phases, to characterise these phases and to determine their solubility, which could control the leaching of spent fuel. One of the main items to carry out the objectives before indicated in both leaching and co-precipitation/solubility experiments is to perform a extensive solid phases characterisation in order to facilitate the understanding of the processes involved. This report is structured in three parts, the first include experimental procedures, characterisation techniques and solid and solution analyses. The second shows the leaching results obtained by considering the effect of pH, complex formation, redox conditions, surface/volume ratio, etc. The third supply the results of the co-precipitation/solubility studies. The conclusions obtained in this work are considered as the start point of going on and more extensive studies on the mechanisms

  16. Shielding properties of protective thin film coatings and blended concrete compositions for high level waste storage packages

    International Nuclear Information System (INIS)

    Fusco, Michael A.; Winfrey, Leigh; Bourham, Mohamed A.

    2016-01-01

    Highlights: • Measured linear attenuation coefficients are the same for bare and coated steels. • Gamma mean free path is much larger than coating thickness; buildup is negligible. • ‘Concrete-6’ reduces exposure rate outside spent fuel cask significantly over ordinary concrete. - Abstract: Various thin film coatings have been proposed to protect stainless steel high level waste (HLW) containers from premature failure due to localized corrosion, hydrogen embrittlement, and mechanical wear. These coatings include TiN, ZrO 2 , MoS 2 , TiO 2 , and Al 2 O 3 , to be deposited either in multiple layers or as a thicker, single-layer composite. Linear attenuation coefficients of these materials have been simulated using MicroShield and measured experimentally for various photon energies. Additionally, spent fuel casks with overpacks made of two different types of concrete were simulated to compare exposure rate at the cask surface. In the energy range that is significant for high level waste storage all coating materials possess very similar attenuation behavior. A specialty concrete, containing magnetite (Fe 3 O 4 ) and lead oxide (PbO), reduces the exposure rate at the outer surface of the overpack by several orders of magnitude. The higher-Z elements not present in ordinary concrete greatly increase attenuation of intermediate-energy gammas (0.4–1.0 MeV). The thin film coatings do not affect the shielding capabilities of the HLW packaging, as their total proposed thickness is nearly three orders of magnitude less than the mean free path (MFP) of the primary photons of interest.

  17. Comparison of the intermediate storage periods and areas required for final storage of high-level radioactive waste and spent fuel in various types of host rock

    International Nuclear Information System (INIS)

    Mueller-Hoeppe, N.; Lerch, C.; Jobmann, M.; Filbert, W.

    2005-01-01

    The present new version of the German concept for radioactive waste and spent fuel management is based on the assumption that a repository for high-level waste and spent fuel will not be required until 2030. One reason frequently given for this date is the intermediate storage period of at least forty years to allow the very high initial heat generation to decay. However, calculations performed by the authors have shown that the minimum intermediate storage period for a repository in rock salt is only between four and nineteen years, depending on the final storage concept and the load of the waste package. In clay as a host rock, the minimum intermediate storage times were calculated to be between 31 and 142 years; the same time spans are expected to apply to final storage in magmatic rock, such as granite. The maximum permissible loads of a container holding spent fuel in salt are many times those in clay and granite, respectively. It was also seen that the area requirement for final storage of the same waste structures is roughly a factor of ten higher in clay than in salt. The differences between granite and salt are similar. The reasons for these grave differences, on the one hand, are the better thermal conductivity of salt and, on the other hand, the better heat tolerance of the crushed salt used as backfill material compared to that of bentonite used in the clay and granite concepts. While salt will allow temperatures of up to 200 C, the maximum temperature in bentonite is limited to 100 C. (orig.)

  18. Status of nuclear fuel reprocessing, spent fuel storage, and high-level waste disposal. Nuclear Fuel Cycle Committee, California Energy Resources Conservation and Development Commission. Draft report

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    An analysis of the current status of technologies and issues in the major portions of the back-end of the nuclear fuel cycle is presented. The discussion on nuclear fuel reprocessing covers the reprocessing requirement, reprocessing technology assessment, technology for operation of reprocessing plants, and approval of reprocessing plants. The chapter devoted to spent fuel storage covers the spent fuel storge problem, the legislative response, options for maintaining full core discharge capacity, prospective availability of alterntive storage options, and the outlook for California. The existence of a demonstrated, developed high-level waste disposal technology is reviewed. Recommendations for Federal programs on high-level waste disposal are made

  19. Ageing management program for the Spanish low and intermediate level waste disposal and spent fuel and high-level waste centralised storage facilities

    Science.gov (United States)

    Zuloaga, P.; Ordoñez, M.; Andrade, C.; Castellote, M.

    2011-04-01

    The generic design of the centralised spent fuel storage facility was approved by the Spanish Safety Authority in 2006. The planned operational life is 60 years, while the design service life is 100 years. Durability studies and surveillance of the behaviour have been considered from the initial design steps, taking into account the accessibility limitations and temperatures involved. The paper presents an overview of the ageing management program set in support of the Performance Assessment and Safety Review of El Cabril low and intermediate level waste (LILW) disposal facility. Based on the experience gained for LILW, ENRESA has developed a preliminary definition of the Ageing Management Plan for the Centralised Interim Storage Facility of spent Fuel and High Level Waste (HLW), which addresses the behaviour of spent fuel, its retrievability, the confinement system and the reinforced concrete structure. It includes tests plans and surveillance design considerations, based on the El Cabril LILW disposal facility.

  20. Ageing management program for the Spanish low and intermediate level waste disposal and spent fuel and high-level waste centralised storage facilities

    Directory of Open Access Journals (Sweden)

    Andrade C.

    2011-04-01

    Full Text Available The generic design of the centralised spent fuel storage facility was approved by the Spanish Safety Authority in 2006. The planned operational life is 60 years, while the design service life is 100 years. Durability studies and surveillance of the behaviour have been considered from the initial design steps, taking into account the accessibility limitations and temperatures involved. The paper presents an overview of the ageing management program set in support of the Performance Assessment and Safety Review of El Cabril low and intermediate level waste (LILW disposal facility. Based on the experience gained for LILW, ENRESA has developed a preliminary definition of the Ageing Management Plan for the Centralised Interim Storage Facility of spent Fuel and High Level Waste (HLW, which addresses the behaviour of spent fuel, its retrievability, the confinement system and the reinforced concrete structure. It includes tests plans and surveillance design considerations, based on the El Cabril LILW disposal facility.

  1. High-level radioactive wastes

    International Nuclear Information System (INIS)

    Grissom, M.C.

    1982-10-01

    This bibliography contains 812 citations on high-level radioactive wastes included in the Department of Energy's Energy Data Base from January 1981 through July 1982. These citations are to research reports, journal articles, books, patents, theses, and conference papers from worldwide sources. Five indexes are provided: Corporate Author, Personal Author, Subject, Contract Number, and Report Number

  2. Use of base isolation techniques for the design of high-level waste storage facility enclosure at INEL

    International Nuclear Information System (INIS)

    Vallenas, J.M.; Wong, Chun K.; Beer, M.J.

    1993-08-01

    Current Department of Energy criteria for facilities subjected to natural hazards provide guidelines to place facilities or portions of facilities into usage categories. Usage categories are based on characteristics such as mission dependence, type of hazardous materials involved, and performance goals. Seismic requirements are significantly more stringent for facilities falling into higher ''hazard facility use categories''. A special problem arises in cases where a facility or portion of a facility is dependent on another facility of lower ''hazard facility use category'' for support or protection. Creative solutions can minimize the cost Unpact of ensuring that the lower category item does not compromise the performance of the higher category item. In this paper, a base isolation solution is provided for a ''low hazard facility use category'' weather enclosure designed so it will not collapse onto a ''high hazard facility use category'' high level waste storage facility at INEL. This solution is compared to other more conventional procedures. Details, practical limitations, licensing and regulatory considerations, and cost comparisons are provided

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  4. Materials for high-level waste containment

    International Nuclear Information System (INIS)

    Marsh, G.P.

    1982-01-01

    The function of the high-level radioactive waste container in storage and of a container/overpack combination in disposal is considered. The consequent properties required from potential fabrication materials are discussed. The strategy adopted in selecting containment materials and the experimental programme underway to evaluate them are described. (U.K.)

  5. High polymer composites for containers for the long-term storage of spent nuclear fuel and high level radioactive waste

    International Nuclear Information System (INIS)

    Bonin, H.W.; Vui, V.T.; Legault, J.-F.

    1997-01-01

    sufficiently shielded from the radiations emitted by the spent nuclear fuel or other high level radioactive waste, this material may well be an interesting candidate for this application. More recent work at RMC on the effects of radiations on PEEK has demonstrated that this polymer thermoplastic material was even superior to epoxies under radiation environments. Part of this research concentrated on the estimation of the doses accumulated in the container wall over the years using three basic models for the container: one without filling material, one with glass beads as proposed by AECL, and one using thorium dioxide (ThO 2 ) as filling material. This choice is based on the excellent physical and chemical properties of this compound (resistance to corrosion in particular) and to the expected low cost since thorium is usually discarded in the tailings of uranium mine concentrating plants. The dose calculations were carried out using the Microshield software and showed that both the epoxy and the PEEK could maintain structural integrity provided that they are shielded sufficiently against the radiations emitted by the high level radioactive waste. This research investigated also the resistance to the mechanical forces to which the container walls would be submitted in the underground vaults and it was concluded that these materials displayed sufficient mechanical strength for such application. It is permitted the identification of several aspects of the design of the storage containers that needed closer investigation. (author)

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

    International Nuclear Information System (INIS)

    Burt, D.L.

    1994-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Burt, D.L.

    1994-04-01

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

  8. Optimizing High Level Waste Disposal

    International Nuclear Information System (INIS)

    Dirk Gombert

    2005-01-01

    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

  9. Disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    Glasby, G.P.

    1977-01-01

    Although controversy surrounding the possible introduction of nuclear power into New Zealand has raised many points including radiation hazards, reactor safety, capital costs, sources of uranium and earthquake risks on the one hand versus energy conservation and alternative sources of energy on the other, one problem remains paramount and is of global significance - the storage and dumping of the high-level radioactive wastes of the reactor core. The generation of abundant supplies of energy now in return for the storage of these long-lived highly radioactive wastes has been dubbed the so-called Faustian bargain. This article discusses the growth of the nuclear industry and its implications to high-level waste disposal particularly in the deep-sea bed. (auth.)

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

    International Nuclear Information System (INIS)

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

    1978-09-01

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

  11. Status of Closure Welding Technology of Canister for Transportation and Storage of High Level Radioactive Material and Waste

    International Nuclear Information System (INIS)

    Lee, H. J.; Bang, K. S.; Seo, K. S.; Seo, C. S.

    2010-10-01

    Closure seal welding is one of the key technologies in fabricating and handling the canister which is used for transportation and storage of high radioactive material and waste. Simple industrial fabrication processes are used before filling the radioactive waste into the canister. But, automatic and remote processes should be used after filling the radioactive material because the thickness of canister is not sufficient to shield the high radiation from filled material or waste. In order to simplify the welding process the closure structure of canister and the sealing method are investigated and developed properly. Two types of radioactive materials such as vitrified waste and compacted solid waste are produced in nuclear industry. Because the filling method of two types of waste is different, the shapes of closure and opening of canister and welding method is also different. The canister shape and sealing method should be standardized to standardize the handling facilities and inspection process such as leak test after closure welding. In order to improve the productivity of disposal and compatibility of the canister, the structure and shape of canister should be standardized considering the type of waste. Two kind of welding process such as arc welding and resistance welding are reported and used in the field. In the arc welding process GTAW and PAW are considered proper processes for closure welding. The closure seal welding process can be selected by considering material of canister, thickness of body, productivity, and applicable codes and rules. Because the storage time of nuclear waste in canister is very long, at least 20 years, the long-time corrosion at the weld should be estimated including mechanical integrity. Recently, the mitigation of residual stress around weld region, which causes stress corrosion cracking, is also interesting research issue

  12. Decommissioning high-level waste surface facilities

    International Nuclear Information System (INIS)

    1978-04-01

    The protective storage, entombment and dismantlement options of decommissioning a High-Level Waste Surface Facility (HLWSF) was investigated. A reference conceptual design for the facility was developed based on the designs of similar facilities. State-of-the-art decommissioning technologies were identified. Program plans and cost estimates for decommissioning the reference conceptual designs were developed. Good engineering design concepts were on the basis of this work identified

  13. National high-level waste systems analysis

    International Nuclear Information System (INIS)

    Kristofferson, K.; O'Holleran, T.P.

    1996-01-01

    Previously, no mechanism existed that provided a systematic, interrelated view or national perspective of all high-level waste treatment and storage systems that the US Department of Energy manages. The impacts of budgetary constraints and repository availability on storage and treatment must be assessed against existing and pending negotiated milestones for their impact on the overall HLW system. This assessment can give DOE a complex-wide view of the availability of waste treatment and help project the time required to prepare HLW for disposal. Facilities, throughputs, schedules, and milestones were modeled to ascertain the treatment and storage systems resource requirements at the Hanford Site, Savannah River Site, Idaho National Engineering Laboratory, and West Valley Demonstration Project. The impacts of various treatment system availabilities on schedule and throughput were compared to repository readiness to determine the prudent application of resources. To assess the various impacts, the model was exercised against a number of plausible scenarios as discussed in this paper

  14. High level waste canister emplacement and retrieval concepts study

    International Nuclear Information System (INIS)

    1975-09-01

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

  15. Other-than-high-level waste

    International Nuclear Information System (INIS)

    Bray, G.R.

    1976-01-01

    The main emphasis of the work in the area of partitioning transuranic elements from waste has been in the area of high-level liquid waste. But there are ''other-than-high-level wastes'' generated by the back end of the nuclear fuel cycle that are both large in volume and contaminated with significant quantities of transuranic elements. The combined volume of these other wastes is approximately 50 times that of the solidified high-level waste. These other wastes also contain up to 75% of the transuranic elements associated with waste generated by the back end of the fuel cycle. Therefore, any detailed evaluation of partitioning as a viable waste management option must address both high-level wastes and ''other-than-high-level wastes.''

  16. Process for solidifying high-level nuclear waste

    Science.gov (United States)

    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.

  17. CASTOR {sup ®} and CONSTOR {sup ®}. A well established system for the dry storage of spent fuel and high level waste

    Energy Technology Data Exchange (ETDEWEB)

    Wimmer, Hannes; Skrzyppek, Juergen; Koebl, Michael [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2015-06-01

    The German company GNS Gesellschaft fuer Nuklear-Service mbH today looks back on more than 30 years of operational experience with dual-purpose casks for the transport and storage of spent nuclear fuel (SNF) from nuclear power plants and high level waste (HLW) from reprocessing. Following customer demands, GNS developed two different cask types for SNF. By now, almost 1,300 GNS-casks are in operation worldwide. This article gives an overview over several national and international projects and shows the bandwidth of customised solutions by GNS.

  18. Chemical decomposition of high-level nuclear waste storage/disposal glasses under irradiation. 1997 annual progress report

    International Nuclear Information System (INIS)

    Griscom, D.L.; Merzbacher, C.I.

    1997-01-01

    'The objective of this research is to use the sensitive technique of electron spin resonance (ESR) to look for evidence of radiation-induced chemical decomposition of vitreous forms contemplated for immobilization of plutonium and/or high-level nuclear wastes, to interpret this evidence in terms of existing knowledge of glass structure, and to recommend certain materials for further study by other techniques, particularly electron microscopy and measurements of gas evolution by high-vacuum mass spectroscopy. Previous ESR studies had demonstrated that an effect of y rays on a simple binary potassium silicate glass was to induce superoxide (O 2 - ) and ozonide (O 3 - ) as relatively stable product of long-term irradiation Accordingly, some of the first experiments performed as a part of the present effort involved repeating this work. A glass of composition 44 K 2 O: 56 SiO 2 was prepared from reagent grade K 2 CO3 and SiO 2 powders melted in a Pt crucible in air at 1,200 C for 1.5 hr. A sample irradiated to a dose of 1 MGy (1 MGy = 10 8 rad) indeed yielded the same ESR results as before. To test the notion that the complex oxygen ions detected may be harbingers of radiation-induced phase separation or bubble formation, a small-angle neutron scattering (SANS) experiment was performed. SANS is theoretically capable of detecting voids or bubbles as small as 10 305 in diameter. A preliminary experiment was carried out with the collaboration of Dr. John Barker (NIST). The SANS spectra for the irradiated and unirradiated samples were indistiguishable. A relatively high incoherent background (probably due to the presence of protons) may obscure scattering from small gas bubbles and therefore decrease the effective resolution of this technique. No further SANS experiments are planned at this time.'

  19. Management of high level radioactive waste

    International Nuclear Information System (INIS)

    Redon, A.; Mamelle, J.; Chambon, M.

    1977-01-01

    The world wide needs in reprocessing will reach the value of 10.000 t/y of irradiated fuels, in the mid of the 80's. Several countries will have planned, in their nuclear programme, the construction of reprocessing plants with a 1500 t/y capacity, corresponding to 50.000 MWe installed. At such a level, the solidification of the radioactive waste will become imperative. For this reason, all efforts, in France, have been directed towards the realization of industrial plants able of solidifying the fission products as a glassy material. The advantages of this decision, and the reasons for it are presented. The continuing development work, and the conditions and methods of storing the high-level wastes prior to solidification, and of the interim storage (for thermal decay) and the ultimate disposal after solidification are described [fr

  20. High-level waste processing and disposal

    International Nuclear Information System (INIS)

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

    1984-11-01

    Without reprocessing, spent LWR fuel itself is generally considered an acceptable waste form. With reprocessing, borosilicate glass canisters, have now gained general acceptance for waste immobilization. The current first choice for disposal is emplacement in an engineered structure in a mined cavern at a depth of 500-1000 meters. A variety of rock types are being investigated including basalt, clay, granite, salt, shale, and volcanic tuff. This paper gives specific coverage to the national high level waste disposal plans for France, the Federal Republic of Germany, Japan and the United States. The French nuclear program assumes prompt reprocessing of its spent fuels, and France has already constructed the AVM. Two larger borosilicate glass plants are planned for a new French reprocessing plant at La Hague. France plans to hold the glass canisters in near-surface storage for a forty to sixty year cooling period and then to place them into a mined repository. The FRG and Japan also plan reprocessing for their LWR fuels. Both are currently having some fuel reprocessed by France, but both are also planning reprocessing plants which will include waste vitrification facilities. West Germany is now constructing the PAMELA Plant at Mol, Belgium to vitrify high level reprocessing wastes at the shutdown Eurochemic Plant. Japan is now operating a vitrification mockup test facility and plans a pilot plant facility at the Tokai reprocessing plant by 1990. Both countries have active geologic repository programs. The United State program assumes little LWR fuel reprocessing and is thus primarily aimed at direct disposal of spent fuel into mined repositories. However, the US have two borosilicate glass plants under construction to vitrify existing reprocessing wastes

  1. High level waste fixation in cermet form

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  2. High-level nuclear waste disposal

    International Nuclear Information System (INIS)

    Burkholder, H.C.

    1985-01-01

    The meeting was timely because many countries had begun their site selection processes and their engineering designs were becoming well-defined. The technology of nuclear waste disposal was maturing, and the institutional issues arising from the implementation of that technology were being confronted. Accordingly, the program was structured to consider both the technical and institutional aspects of the subject. The meeting started with a review of the status of the disposal programs in eight countries and three international nuclear waste management organizations. These invited presentations allowed listeners to understand the similarities and differences among the various national approaches to solving this very international problem. Then seven invited presentations describing nuclear waste disposal from different perspectives were made. These included: legal and judicial, electric utility, state governor, ethical, and technical perspectives. These invited presentations uncovered several issues that may need to be resolved before high-level nuclear wastes can be emplaced in a geologic repository in the United States. Finally, there were sixty-six contributed technical presentations organized in ten sessions around six general topics: site characterization and selection, repository design and in-situ testing, package design and testing, disposal system performance, disposal and storage system cost, and disposal in the overall waste management system context. These contributed presentations provided listeners with the results of recent applied RandD in each of the subject areas

  3. National high-level waste systems analysis report

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  4. National high-level waste systems analysis report

    International Nuclear Information System (INIS)

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

    1995-09-01

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

  5. Properties and characteristics of high-level waste glass

    International Nuclear Information System (INIS)

    Ross, W.A.

    1977-01-01

    This paper has briefly reviewed many of the characteristics and properties of high-level waste glasses. From this review, it can be noted that glass has many desirable properties for solidification of high-level wastes. The most important of these include: (1) its low leach rate; (2) the ability to tolerate large changes in waste composition; (3) the tolerance of anticipated storage temperatures; (4) its low surface area even after thermal shock or impact

  6. Selection of concepts for monitored retrievable storage of spent nuclear fuel and high-level radioactive wastes

    International Nuclear Information System (INIS)

    1984-04-01

    The monitored retrievable storage (MRS) concepts considered are: metal cask (stationary and transportable); concrete cask (sealed storage cask); concrete cask-in-trench; field drywell; tunnel drywell; open cycle vault; closed cycle vault; and tunnel rack vault. These concepts were compared primarily upon the relative performance of the storage units on seven criteria which together encompass the key considerations for selecting an MRS concept, namely their ability to satisfy the MRS mission requirements. These criteria were: safety and licensing; environmental impacts; socioeconomic impacts; siting requirements; cost; concept maturity; and flexibility. Evaluations of the candidate concepts indicate that all of the concepts could satisfactorily serve in an MRS facility. However, using the above criteria, the two concepts selected for further design studies are the concrete cask (primary concept), better entitled the sealed storage cask, and the field drywell (alternate concept). It was recognized that the transportable metal storage cask may be used to supplement at-reactor storage until such time as the repository or MRS becomes available. Consequently, a hybrid storage facility may be required (e.g., one using concrete casks or field drywells, with the capability of receiving and storing the transportable cask). Both the concrete cask and the field drywell concepts can easily accommodate the transportable cask. Further design efforts will ensure the compatibility of the MRS designs with the transportable cask

  7. Evaluation of concepts for monitored retrievable storage of spent nuclear fuel and high-level radioactive waste

    International Nuclear Information System (INIS)

    Triplett, M.B.; Smith, R.I.

    1984-04-01

    The primary mission selected by DOE for the monitored retrieval storage (MRS) system is to provide an alternative means of storage in the event that the repository program is delayed. The MRS concepts considered were the eight concepts included in the MRS Research and Development Report to Congress (DOE 1983). These concepts are: metal cask (stationary and transportable); concrete cask (sealed storage cask); concrete cask-in-trench; field drywell; tunnel drywell; open cycle vault; closed cycle vault; and tunnel rack vault. Conceptual design analyses were performed for the candidate concepts using a common set of design requirements specified in consideration of the MRS mission

  8. On the pathway towards disposal. The need for long-term interim storage of high-level nuclear waste

    International Nuclear Information System (INIS)

    Budelmann, Harald; Koehnke, Dennis; Reichardt, Manuel

    2017-01-01

    The disposal of spent nuclear fuel is a still unsolved problem with social, ethical, economical, ecological and political dimensions. The stagnating decision process on the final repository concept in several countries has the consequence of the inclusion of long-term interim storage into the disposal concept. The contribution discusses several approaches. This opens the question whether the long-term interim storage is a matter of delaying tactic or a pragmatic solution on the way to a final repository.

  9. Current high-level waste solidification technology

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  10. French high level wastes management

    International Nuclear Information System (INIS)

    Gauvenet, A.J.; Sombret, C.G.

    1980-06-01

    The first French spent fuel reprocessing plant went on stream in 1956 at Marcoule. Since then, all French irradiated fuels and some foreign spent fuels have been reprocessed either in this plant or in a subsequent plant built at La Hague. Marcoule is primarily devoted to metallic fuels, and La Hague to oxide fuels. The fission products solutions generated by reprocessing are acid liquids. They are stored on site in double walled stainless steel tanks fitted with a cooling device to deal with thermal release due to radioactive decay. Although these liquids are retrievable and can be transfered from one tank to another, and in spite of the fact that no disturbance such as overheating or leakage has ever occurred, a decision was made to solidify these solutions in order to make interim storage and, later on, ultimate disposal, safer and easier to control. Glass was chosen, because it is a flexible medium to deal with and it answers the quality requirements of ultimate disposal as well as the manufacturing constraints, such as equipment corrosion, volatilization during fabrication, and suitability to casting into canisters

  11. Spent fuel test-climax: a test of geologic storage of high-level waste in granite

    International Nuclear Information System (INIS)

    Ramspott, L.D.; Ballou, L.B.; Patrick, W.C.

    1981-01-01

    A test of retrievable geologic storage of spent fuel assemblies from an operating commercial nuclear reactor is underway at the Nevada Test Site (NTS) of the US Department of Energy. This generic test is located 420 m below the surface in the Climax granitic stock. Eleven canisters of spent fuel approximately 2.5 years out of reactor core (about 1.6 kW/canister thermal output) were emplaced in a storage drift along with 6 electrical simulator canisters. Two adjacent drifts contain electrical heaters, which are operated to simulate within the test array the thermal field of a large repository. Fuel was loaded during April to May 1980 and initial results of the test will be presented

  12. Brief draft on surface and subsurface storage of high level and long-lived radioactive wastes. Spent fuels synthesis file

    International Nuclear Information System (INIS)

    Dumas, C.; Jaecki, P.

    2002-01-01

    This document makes a synthesis of the results of two brief draft studies performed in 2002 about the surface and subsurface storage of spent fuels. These studies stress on the long duration aspect of the disposal: feasibility of a secular disposal facility, potential risks and safety level of such a facility, estimation of the initial investment and of operation and maintenance costs. The main points of the specifications and the input data are presented first, and then the subsurface and surface draft studies are described. Content: specifications (imposed design principles and options, dry corrosion, input data); subsurface storage (description and design options, thermal dimensioning and ventilation, geotechnical stability of the facility, subsurface water management, dry corrosion, infrastructure durability, safety, monitoring, security and physical protection, technical-economical aspects, case of Mox fuel, case of glass packages); surface storage (description and design options, thermal dimensioning and ventilation, mechanical dimensioning of the facility, dry corrosion, infrastructure durability, safety, monitoring, security and physical protection, technical-economical aspects, case of Mox fuel, case of glass packages); conclusions and perspectives. (J.S.)

  13. Identification and understanding the factors affecting the public and political acceptance of long term storage of spent fuel and high-level radioactive wastes

    International Nuclear Information System (INIS)

    Gorea, Valica

    2006-01-01

    , during the storage. There is a powerful scientific and technical consensus according to which a well placed final repository must be provided with some barriers, including the geological and engineered ones. The wide-spread and accepted concept of final repository for spent fuel and high level waste is the so-called geological disposal which means the solid radioactive waste storage in underground repositories in a stable geological structure (ordinary at some hundreds of meters deep) so ensuring the isolation of radionuclides from biosphere on long term. The types of radioactive waste (high level and long life waste) which will be placed in geological repository are established by the national strategy and politics for the safe management of the radioactive waste. The development and implementation process of a repository can be achieved in the next non compulsory steps: - Developing the national politics in the field of safe management of the radioactive waste - Developing the legislative and institutional framework - Developing the technical concept of geological repository - Initiation of underground and surface investigations for the characterization of the host rock - Selecting the suitable site for an underground repository - Design, licensing for construction (which takes into account the environmental impact, nuclear safety and so on), operation and shut down. The decision to construct a geological repository must be taken by the Government or by the producers of waste. The steps for the establishing the legal framework, regulation and licensing are prerogatives of the Government, all the others steps could be achieved by non governmental organizations. The achievement of a geological repository, from the technical concept to the operation may last even more than 50 years and its operation some hundreds years. The paper contains the following sections: 1. Introduction; 2. Public and political acceptance; 3. Factors which affect the public and political

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

    International Nuclear Information System (INIS)

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

    1976-06-01

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

  15. Answers to your questions on high-level nuclear waste

    International Nuclear Information System (INIS)

    1987-11-01

    This booklet contains answers to frequently asked questions about high-level nuclear wastes. Written for the layperson, the document contains basic information on the hazards of radiation, the Nuclear Waste Management Program, the proposed geologic repository, the proposed monitored retrievable storage facility, risk assessment, and public participation in the program

  16. High-level radioactive waste management

    International Nuclear Information System (INIS)

    Schneider, K.J.; Liikala, R.C.

    1974-01-01

    High-level radioactive waste in the U.S. will be converted to an encapsulated solid and shipped to a Federal repository for retrievable storage for extended periods. Meanwhile the development of concepts for ultimate disposal of the waste which the Federal Government would manage is being actively pursued. A number of promising concepts have been proposed, for which there is high confidence that one or more will be suitable for long-term, ultimate disposal. Initial evaluations of technical (or theoretical) feasibility for the various waste disposal concepts show that in the broad category, (i.e., geologic, seabed, ice sheet, extraterrestrial, and transmutation) all meet the criteria for judging feasibility, though a few alternatives within these categories do not. Preliminary cost estimates show that, although many millions of dollars may be required, the cost for even the most exotic concepts is small relative to the total cost of electric power generation. For example, the cost estimates for terrestrial disposal concepts are less than 1 percent of the total generating costs. The cost for actinide transmutation is estimated at around 1 percent of generation costs, while actinide element disposal in space is less than 5 percent of generating costs. Thus neither technical feasibility nor cost seems to be a no-go factor in selecting a waste management system. The seabed, ice sheet, and space disposal concepts face international policy constraints. The information being developed currently in safety, environmental concern, and public response will be important factors in determining which concepts appear most promising for further development

  17. Implications of monitored retrievable storage for geologic disposal of spent nuclear fuel and high level radioactive waste

    International Nuclear Information System (INIS)

    Halstead, R.J.; Kidwell, S.M.; Woodbury, D.

    1986-01-01

    The integral monitored retrievable storage (I-MRS) proposal has major implications for geologic disposal. This paper reviews the positive and negative implications from the standpoint of a potential repository host state. Recommendations for improving the I-MRS proposal include: eliminate provisions restricting I-MRS backup role; add provisions to prevent I-MRS from becoming a permanent disposal facility; optimize reactor-to-I-MRS transportation system; further shift preclosure operations from repository to I-MRS; defer decision on rod consolidation; repeat the I-MRS site selection process; eliminate any potential linkage between I-MRS and nuclear weapons programs; and incorporate I-MRS in the repository siting program

  18. Vitrification of high level wastes in France

    International Nuclear Information System (INIS)

    Sombret, C.

    1984-02-01

    A brief historical background of the research and development work conducted in France over 25 years is first presented. Then, the papers deals with the vitrification at (1) the UP1 reprocessing plant (Marcoule) and (2) the UP2 and UP3 reprocessing plants (La Hague). 1) The properties of glass required for high-level radioactive waste vitrification are recalled. The vitrification process and facility of Marcoule are presented. (2) The average characteristics (chemical composition, activity) of LWR fission product solution are given. The glass formulations developed to solidify LWR waste solution must meet the same requirements as those used in the UP1 facility at Marcoule. Three important aspects must be considered with respect to the glass fabrication process: corrosiveness of the molten glass with regard to metals, viscosity of the molten glass, and, volatization during glass fabrication. The glass properties required in view of interim storage and long-term disposal are then largely developed. Two identical vitrification facilities are planned for the site: T7, to process the UP2 throughput, and T7 for the UP3 plant. A prototype unit was built and operated at Marcoule

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

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

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

    International Nuclear Information System (INIS)

    1994-04-01

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

  1. CASTOR {sup registered} HAW28M - a high heat load cask for transport and storage of vitrified high level waste containers

    Energy Technology Data Exchange (ETDEWEB)

    Vossnacke, A.; Klein, K.; Kuehne, B. [GNS Gesellschaft fuer Nuklear-Service mbH/GNB, Essen (Germany)

    2004-07-01

    Within the German return programme for vitrified high level waste (HLW) from reprocessing at COGEMA and BNFL up to now 39 casks loaded with 28 containers each were transported back to Germany and are stored in the Interim Storage Facility Gorleben (TBL-G) for up to 40 years. For transport and storage in all but one case the GNB casks CASTOR {sup registered} HAW 20/28 CG have been used. This cask type is designed to accommodate 20 or 28 HLW containers with a total thermal power of 45 kW maximum. In the near future, among the high level waste, which has to be returned to Germany, there will be an increasing number of containers of which the heat capacity and radioactive inventory will exceed the technical limits of the CASTOR {sup registered} HAW 20/28 CG. Therefore GNB has started the development of a new cask generation, named CASTOR {sup registered} HAW28M, meeting these future requirements. The CASTOR {sup registered} HAW28M is especially developed for the transport of vitrified residues from France and Great Britain to Germany. It complies with the international regulations for type B packages according to IAEA (International Atomic Energy Agency). It is thus guaranteed that even in case of any accident the cask body and the lid system remain functional and the safe confinement of the radioactive contents remains intact during transport. The CASTOR {sup registered} HAW28M fulfills not only the requirements for transport but also the acceptance criteria of interim storage: radiation shielding, heat dissipation, safe confinement under both normal and hypothetical accident conditions. Storage buildings such as the TBL-G simply support the safety functions of the cask. The challenge for the development results from higher requirements of the technical specification, particularly related to fuel which is reprocessed. As a consequence of the reprocessing of fuel with increased enrichment and burn up, higher heat capacity and sophisticated shielding measures have to be

  2. PAIRWISE BLENDING OF HIGH LEVEL WASTE

    International Nuclear Information System (INIS)

    CERTA, P.J.

    2006-01-01

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

  3. Cermets for high level waste containment

    International Nuclear Information System (INIS)

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

    1978-01-01

    Cermet materials are currently under investigation as an alternate for the primary containment of high level wastes. The cermet in this study is an iron--nickel base metal matrix containing uniformly dispersed, micron-size fission product oxides, aluminosilicates, and titanates. Cermets possess high thermal conductivity, and typical waste loading of 70 wt % with volume reduction factors of 2 to 200 and low processing volatility losses have been realized. Preliminary leach studies indicate a leach resistance comparable to other candidate waste forms; however, more quantitative data are required. Actual waste studies have begun on NFS Acid Thorex, SRP dried sludge and fresh, unneutralized SRP process wastes

  4. Managing the nation's commercial high-level radioactive waste

    International Nuclear Information System (INIS)

    1985-03-01

    This report presents the findings and conclusions of OTA's analysis of Federal policy for the management of commercial high-level radioactive waste. It represents a major update and expansion of the Analysis presented to Congress in our summary report, Managing Commercial High-Level Radioactive Waste, published in April of 1982 (NWPA). This new report is intended to contribute to the implementation of NWPA, and in particular to Congressional review of three major documents that DOE will submit to the 99th Congress: a Mission Plan for the waste management program; a monitored retrievable storage (MRS) proposal; and a report on mechanisms for financing and managing the waste program. The assessment was originally focused on the ocean disposal of nuclear waste. OTA later broadened the study to include all aspects of high-level waste disposal. The major findings of the original analysis were published in OTA's 1982 summary report

  5. Overview: Defense high-level waste technology program

    International Nuclear Information System (INIS)

    Shupe, M.W.; Turner, D.A.

    1987-01-01

    Defense high-level waste generated by atomic energy defense activities is stored on an interim basis at three U.S. Department of Energy (DOE) operating locations; the Savannah River Plant in South Carolina, the Hanford Site in Washington, and the Idaho National Engineering Laboratory in Idaho. Responsibility for the permanent disposal of this waste resides with DOE's Office of Defense Waste and Transportation Management. The objective of the Defense High-Level Wast Technology Program is to develop the technology for ending interim storage and achieving permanent disposal of all U.S. defense high-level waste. New and readily retrievable high-level waste are immobilized for disposal in a geologic repository. Other high-level waste will be stabilized in-place if, after completion of the National Environmental Policy Act (NEPA) process, it is determined, on a site-specific basis, that this option is safe, cost effective and environmentally sound. The immediate program focus is on implementing the waste disposal strategy selected in compliance with the NEPA process at Savannah River, while continuing progress toward development of final waste disposal strategies at Hanford and Idaho. This paper presents an overview of the technology development program which supports these waste management activities and an assessment of the impact that recent and anticipated legal and institutional developments are expected to have on the program

  6. High-Level Waste Melter Study Report

    Energy Technology Data Exchange (ETDEWEB)

    Perez, 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.

  7. High-level waste processing and disposal

    International Nuclear Information System (INIS)

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

    1984-01-01

    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 add about 5 to 10% to the costs of nuclear electric power. The third conclusion is less optimistic. Political problems remain formidable in highly conservative regulations, in qualifying a final disposal site, and in securing acceptable transport routes

  8. The relationship between the Resource Conservation and Recovery Act and the storage and disposal of spent nuclear fuel and high-level waste

    International Nuclear Information System (INIS)

    Gertz, C.P.; Cloke, P.L.

    1993-01-01

    This paper addresses the potential applicability of the requirements of the Resource Conservation and Recovery Act (RCRA) to the disposal of spent commercial nuclear fuel and of high-level (vitrified) radioactive waste. The Atomic Energy Act of 1954, as amended, and the associated regulations issued by the US NRC provides many requirements that apply to these waste forms and largely, if not entirely, pre-empts the applicability of RCRA. The RCRA would apply only to the non-radioactive components of these wastes, and then only in respect to hazardous components. In view of these restrictions it becomes important to evaluate whether any components of spent fuel or high-level waste are toxic, as defined by the RCRA regulations. Present indications are that they are not and, hence, the US DOE is proceeding on the basis that these wastes and others that may be generated in the future are non-hazardous in respect to RCRA definitions

  9. High-level radioactive wastes. Supplement 1

    International Nuclear Information System (INIS)

    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. Overview of high-level waste management accomplishments

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  11. Timing of High-level Waste Disposal

    International Nuclear Information System (INIS)

    2008-01-01

    This study identifies key factors influencing the timing of high-level waste (HLW) disposal and examines how social acceptability, technical soundness, environmental responsibility and economic feasibility impact on national strategies for HLW management and disposal. Based on case study analyses, it also presents the strategic approaches adopted in a number of national policies to address public concerns and civil society requirements regarding long-term stewardship of high-level radioactive waste. The findings and conclusions of the study confirm the importance of informing all stakeholders and involving them in the decision-making process in order to implement HLW disposal strategies successfully. This study will be of considerable interest to nuclear energy policy makers and analysts as well as to experts in the area of radioactive waste management and disposal. (author)

  12. Site suitability criteria for solidified high level waste repositories

    International Nuclear Information System (INIS)

    Heckman, R.A.; Holdsworth, T.; Towse, D.F.

    1979-01-01

    Activities devoted to development of regulations, criteria, and standards for storage of solidified high-level radioactive wastes are reported. The work is summarized in sections on site suitability regulations, risk calculations, geological models, aquifer models, human usage model, climatology model, and repository characteristics. Proposed additional analytical work is also summarized

  13. High-Level Waste Vitrification Facility Feasibility Study

    International Nuclear Information System (INIS)

    D. A. Lopez

    1999-01-01

    A ''Settlement Agreement'' between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste now stored at the Idaho Nuclear Technology and Engineering Center will be treated so that it is ready to be moved out of Idaho for disposal by a compliance date of 2035. This report investigates vitrification treatment of the high-level waste in a High-Level Waste Vitrification Facility based on the assumption that no more New Waste Calcining Facility campaigns will be conducted after June 2000. Under this option, the sodium-bearing waste remaining in the Idaho Nuclear Technology and Engineering Center Tank Farm, and newly generated liquid waste produced between now and the start of 2013, will be processed using a different option, such as a Cesium Ion Exchange Facility. The cesium-saturated waste from this other option will be sent to the Calcine Solids Storage Facilities to be mixed with existing calcine. The calcine and cesium-saturated waste will be processed in the High-Level Waste Vitrification Facility by the end of calendar year 2035. In addition, the High-Level Waste Vitrification Facility will process all newly-generated liquid waste produced between 2013 and the end of 2035. Vitrification of this waste is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the waste and pouring it into stainless-steel canisters that will be ready for shipment out of Idaho to a disposal facility by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory until they are sent to a national geologic repository. The operating period for vitrification treatment will be from the end of 2015 through 2035

  14. High-Level Waste Vitrification Facility Feasibility Study

    Energy Technology Data Exchange (ETDEWEB)

    D. A. Lopez

    1999-08-01

    A ''Settlement Agreement'' between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste now stored at the Idaho Nuclear Technology and Engineering Center will be treated so that it is ready to be moved out of Idaho for disposal by a compliance date of 2035. This report investigates vitrification treatment of the high-level waste in a High-Level Waste Vitrification Facility based on the assumption that no more New Waste Calcining Facility campaigns will be conducted after June 2000. Under this option, the sodium-bearing waste remaining in the Idaho Nuclear Technology and Engineering Center Tank Farm, and newly generated liquid waste produced between now and the start of 2013, will be processed using a different option, such as a Cesium Ion Exchange Facility. The cesium-saturated waste from this other option will be sent to the Calcine Solids Storage Facilities to be mixed with existing calcine. The calcine and cesium-saturated waste will be processed in the High-Level Waste Vitrification Facility by the end of calendar year 2035. In addition, the High-Level Waste Vitrification Facility will process all newly-generated liquid waste produced between 2013 and the end of 2035. Vitrification of this waste is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the waste and pouring it into stainless-steel canisters that will be ready for shipment out of Idaho to a disposal facility by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory until they are sent to a national geologic repository. The operating period for vitrification treatment will be from the end of 2015 through 2035.

  15. Design concepts of definitive disposal for high level radioactive wastes

    International Nuclear Information System (INIS)

    Badillo A, V.E.; Alonso V, G.

    2007-01-01

    It is excessively known the importance about finding a solution for the handling and disposition of radioactive waste of all level. However, the polemic is centered in the administration of high level radioactive waste and the worn out fuel, forgetting that the more important volumes of waste its are generated in the categories of low level wastes or of very low level. Depending on the waste that will be confined and of the costs, several technological modalities of definitive disposition exist, in function of the depth of the confinement. The concept of deep geologic storage, technological option proposed more than 40 years ago, it is a concept of isolation of waste of long half life placed in a deep underground installation dug in geologic formations that are characterized by their high stability and their low flow of underground water. In the last decades, they have registered countless progresses in technical and scientific aspects of the geologic storage, making it a reliable technical solution supported with many years of scientific work carried out by numerous institutions in the entire world. In this work the design concepts that apply some countries for the high level waste disposal that its liberate heat are revised and the different geologic formations that have been considered for the storage of this type of wastes. (Author)

  16. International high-level radioactive waste repositories

    International Nuclear Information System (INIS)

    Lin, W.

    1996-01-01

    Although nuclear technologies benefit everyone, the associated nuclear wastes are a widespread and rapidly growing problem. Nuclear power plants are in operation in 25 countries, and are under construction in others. Developing countries are hungry for electricity to promote economic growth; industrialized countries are eager to export nuclear technologies and equipment. These two ingredients, combined with the rapid shrinkage of worldwide fossil fuel reserves, will increase the utilization of nuclear power. All countries utilizing nuclear power produce at least a few tens of tons of spent fuel per year. That spent fuel (and reprocessing products, if any) constitutes high-level nuclear waste. Toxicity, long half-life, and immunity to chemical degradation make such waste an almost permanent threat to human beings. This report discusses the advantages of utilizing repositories for disposal of nuclear wastes

  17. Ramifications of defining high-level waste

    International Nuclear Information System (INIS)

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

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Baskin, K.P.

    1992-01-01

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

  19. Technetium Chemistry in High-Level Waste

    International Nuclear Information System (INIS)

    Hess, Nancy J.

    2006-01-01

    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

  20. Loading, transport and storage of casks of the type CASTOR registered HAW28M in the frame of vitrified high-level waste repatriation from France

    International Nuclear Information System (INIS)

    Horn, Thomas; Graf, Wilhelm; Gosch-Warning, Michaela

    2011-01-01

    Until 2005 the German nuclear power plant operators have contracts with AREVA NC (former COGEMA) and NDA (former BNFL) concerning the reprocessing of spent fuel elements. The reprocessed and vitrified radioactive waste has to be repatriated to Germany. Due to the reprocessing of spent fuel elements with increased burnup and the repatriation after shorter cooling time the total activity and the Cm-244 content of the high-level-waste coquilles have increased since 2008. Consequently the heat output has increased to 2 kW/coquille. Therefore the new transport cask type CASTOR registered HAW28M was developed. The authors describe the design of the casks, the licensing according to the German transport regulations, loading procedures, radiation measurements and shipment completion. In autumn 2011 the repatriation of vitrified high-level waste from France is supposed to be completed with the transport of eleven CASTOR registered HAW28M.

  1. High-level radioactive wastes. Supplement 1

    Energy Technology Data Exchange (ETDEWEB)

    McLaren, L.H. (ed.)

    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.

  2. Remote automatic plasma arc-closure welding of a dry-storage canister for spent nuclear fuel and high-level radioactive waste

    International Nuclear Information System (INIS)

    Sprecace, R.P.; Blankenship, W.P.

    1982-01-01

    A carbon steel storage canister has been designed for the dry encapsulation of spent nuclear fuel assemblies or of logs of vitrified high level radioactive waste. The canister design is in conformance with the requirements of the ASME Code, Section III, Division 1 for a Class 3 vessel. The canisters will be loaded and sealed as part of a completely remote process sequence to be performed in the hot bay of an experimental encapsulation facility at the Nevada Test Site. The final closure to be made is a full penetration butt weld between the canister body, a 12.75-in O.D. x 0.25-in wall pipe, and a mating semiellipsoidal closure lid. Due to a combination of design, application and facility constraints, the closure weld must be made in the 2G position (canister vertical). The plasma arc welding system is described, and the final welding procedure is described and discussed in detail. Several aspects and results of the procedure development activity, which are of both specific and general interest, are highlighted; these include: The critical welding torch features which must be exactly controlled to permit reproducible energy input to, and gas stream interaction with, the weld puddle. A comparison of results using automatic arc voltage control with those obtained using a mechanically fixed initial arc gap. The optimization of a keyhole initiation procedure. A comparison of results using an autogenous keyhole closure procedure with those obtained using a filler metal addition. The sensitivity of the welding process and procedure to variations in joint configuration and dimensions and to variations in base metal chemistry. Finally, the advantages and disadvantages of the plasma arc process for this application are summarized from the current viewpoint, and the applicability of this process to other similar applications is briefly indicated

  3. High-level waste canister envelope study: structural analysis

    International Nuclear Information System (INIS)

    1977-11-01

    The structural integrity of waste canisters, fabricated from standard weight Type 304L stainless steel pipe, was analyzed for sizes ranging from 8 to 24 in. diameter and 10 to 16 feet long under normal, abnormal, and improbable life cycle loading conditions. The canisters are assumed to be filled with vitrified high-level nuclear waste, stored temporarily at a fuel reprocessing plant, and then transported for storage in an underground salt bed or other geologic storage. In each of the three impact conditions studies, the resulting impact force is far greater than the elastic limit capacity of the material. Recommendations are made for further study

  4. Radiolytic bubble formation and level changes in simulated high-level waste salts and sludges -- application to Savannah River Site and Hanford Storage tanks

    International Nuclear Information System (INIS)

    Walker, D.D.; Crawford, C.L.; Bibler, N.E.

    1993-01-01

    Radiolytically-produced bubbles of trapped gas are observed in simulated high-level waste (HLW) damp salt cake exposed to Co-60 gamma radiation. As the damp salt cake is irradiated, its volume increases due to the formation of trapped gas bubbles. Based on the increase in volume, the rate of trapped gas generation varies between 0.04 and 0.2 molecules/100 eV of energy deposited in the damp salt cake. The maximum volume of trapped gas observed in experiments is in the range 21--26 vol %. After reaching these volumes, the gas bubbles begin to escape. The generated gas includes hydrogen, oxygen, and nitrous oxide. The ratio in which these components are produced depends on the composition of the waste. Nitrous oxide production increases with the amount of sodium nitrite. Gases trapped by this mechanism may account for some of the observed level changes in Savannah River Site and Hanford waste tanks

  5. Nuclear waste storage container with metal matrix

    International Nuclear Information System (INIS)

    Sump, K.R.

    1978-01-01

    The invention relates to a storage container for high-level waste having a metal matrix for the high-level waste, thereby providing greater impact strength for the waste container and increasing heat transfer properties

  6. Nuclear waste storage container with metal matrix

    Science.gov (United States)

    Sump, Kenneth R.

    1978-01-01

    The invention relates to a storage container for high-level waste having a metal matrix for the high-level waste, thereby providing greater impact strength for the waste container and increasing heat transfer properties.

  7. Criteria for high-level waste disposal

    International Nuclear Information System (INIS)

    Sousselier, Y.

    1981-01-01

    Disposal of radioactive wastes is storage without the intention of retrieval. But in such storage, it may be useful and in some cases necessary to have the possibility of retrieval at least for a certain period of time. In order to propose some criteria for HLW disposal, one has to examine how this basic concept is to be applied. HLW is waste separated as a raffinate in the first cycle of solvent extraction in reprocessing. Such waste contains the bulk of fission products which have long half lives, therefore the safety of a disposal site, at least after a certain period of time, must be intrinsic, i.e. not based on human intervention. There is a consensus that such a disposal is feasible in a suitable geological formation in which the integrity of the container will be reinforced by several additional barriers. Criteria for disposal can be proposed for all aspects of the question. The author discusses the aims of the safety analysis, particularly the length of time for this analysis, and the acceptable dose commitments resulting from the release of radionuclides, the number and role of each barrier, and a holistic analysis of safety external factors. (Auth.)

  8. Processing vessel for high level radioactive wastes

    International Nuclear Information System (INIS)

    Maekawa, Hiromichi

    1998-01-01

    Upon transferring an overpack having canisters containing high level radioactive wastes sealed therein and burying it into an underground processing hole, an outer shell vessel comprising a steel plate to be fit and contained in the processing hole is formed. A bury-back layer made of dug earth and sand which had been discharged upon forming the processing hole is formed on the inner circumferential wall of the outer shell vessel. A buffer layer having a predetermined thickness is formed on the inner side of the bury-back layer, and the overpack is contained in the hollow portion surrounded by the layer. The opened upper portion of the hollow portion is covered with the buffer layer and the bury-back layer. Since the processing vessel having a shielding performance previously formed on the ground, the state of packing can be observed. In addition, since an operator can directly operates upon transportation and burying of the high level radioactive wastes, remote control is no more necessary. (T.M.)

  9. Managing the nation's commercial high-level radioactive waste

    International Nuclear Information System (INIS)

    Cotton, T.

    1985-01-01

    With the passage of the Nuclear Waste Policy Act of 1982 (NWPA), Congress for the first time established in law a comprehensive Federal policy for commercial high-level radioactive waste management, including interim storage and permanent disposal. NWPA provides sufficient authority for developing and operating a high-level radioactive waste management system based on disposal in mined geologic repositories. Authorization for other types of waste facilities will not be required unless major problems with geologic disposal are discovered, and studies to date have identified no insurmountable technical obstacles to developing geologic repositories. The NWPA requires the Department of Energy (DOE) to submit to Congress three key documents: (1) a Mission Plan, containing both a waste management plan with a schedule for transferring waste to Federal facilities and an implementation program for choosing sites and developing technologies to carry out that plan; (2) a monitored retrievable storage (MRS) proposal, to include a site-specific design for a long-term federal storage facility, an evaluation of whether such an MRS facility is needed and feasible, and an analysis of how an MRS facility would be integrated with the repository program if authorized by Congress; and (3) a study of alternative institutional mechanisms for financing and managing the radioactive waste system, including the option of establishing an independent waste management organization outside of DOE. The Mission Plan and the report on alternative institutional mechanisms were submitted to the 99th US Congress in 1985. The MRS proposal is to be submitted in early 1986. Each of these documents is discussed following an overview of the Nuclear Waste Policy Act of 1982

  10. Status of the French nuclear high level waste disposal

    International Nuclear Information System (INIS)

    Sombret, C.

    1985-09-01

    French research on high level waste processing has led to the development of industrial vitrification facilities. Borosilicate glass is still being investigated for its long-term storage properties, since it is itself a component of the containment system. The other constituents of this system, the engineered barriers, are also being actively investigated. The geological barrier is now being assessed using a methodology applicable to various types of geological formations, and final site qualification should be possible before the end of 1992

  11. Solidification of Savannah River Plant high level waste

    International Nuclear Information System (INIS)

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

    1981-11-01

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

  12. Materials Science of High-Level Nuclear Waste Immobilization

    International Nuclear Information System (INIS)

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

    2009-01-01

    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

  13. Vitrification of high-level liquid wastes

    International Nuclear Information System (INIS)

    Varani, J.L.; Petraitis, E.J.; Vazquez, Antonio.

    1987-01-01

    High-level radioactive liquid wastes produced in the fuel elements reprocessing require, for their disposal, a preliminary treatment by which, through a series of engineering barriers, the dispersion into the biosphere is delayed by 10 000 years. Four groups of compounds are distinguished among a great variety of final products and methods of elaboration. From these, the borosilicate glasses were chosen. Vitrification experiences were made at a laboratory scale with simulated radioactive wastes, employing different compositions of borosilicate glass. The installations are described. A series of tests were carried out on four basic formulae using always the same methodology, consisting of a dry mixture of the vitreous matrix's products and a dry simulated mixture. Several quality tests of the glasses were made 1: Behaviour in leaching following the DIN 12 111 standard; 2: Mechanical resistance; parameters related with the facility of the different glasses for increasing their surface were studied; 3: Degree of devitrification: it is shown that devitrification turns the glasses containing radioactive wastes easily leachable. From all the glasses tested, the composition SiO 2 , Al 2 O 3 , B 2 O 3 , Na 2 O, CaO shows the best retention characteristics. (M.E.L.) [es

  14. Ocean disposal of high level radioactive waste

    International Nuclear Information System (INIS)

    1983-01-01

    This study confirms, subject to limitations of current knowledge, the engineering feasibility of free fall penetrators for High Level Radioactive Waste disposal in deep ocean seabed sediments. Restricted sediment property information is presently the principal bar to an unqualified statement of feasibility. A 10m minimum embedment and a 500 year engineered barrier waste containment life are identified as appropriate basic penetrator design criteria at this stage. A range of designs are considered in which the length, weight and cross section of the penetrator are varied. Penetrators from 3m to 20m long and 2t to 100t in weight constructed of material types and thicknesses to give a 500 year containment life are evaluated. The report concludes that the greatest degree of confidence is associated with performance predictions for 75 to 200 mm thick soft iron and welded joints. A range of lengths and capacities from a 3m long single waste canister penetrator to a 20m long 12 canister design are identified as meriting further study. Estimated embedment depths for this range of penetrator designs lie between 12m and 90m. Alternative manufacture, transport and launch operations are assessed and recommendations are made. (author)

  15. Studies of corrosion in metallic container for storage of high level radioactive wastes; Estudios de corrosion de materiales metalicos para capsulas de almacenamiento de residuos de alta actividad

    Energy Technology Data Exchange (ETDEWEB)

    Azkarate, I; Madina, V; Insausti, M

    1999-11-01

    The metallic container is one of the most important barriers that, along with engineered and natural barriers, will isolate high level nuclear waste in saline and granite geological formations from the geosphere. However, general and localized corrosion modes such as stress corrosion cracking (SCC), pitting, crevice corrosion and hydrogen damage can be active under disposal conditions, so the corrosion behaviour of the metal container material must be carefully studied. Several metals and their alloys have been proposed for the fabrication of nuclear waste containers including carbon steels, stainless steels, titanium and titanium alloys and copper and copper-base alloys. Carbon steels and copper alloys are considered for the two rock formations, titanium is considered for salt environments and the stainless steel only in the case of a granite formation. (Author)

  16. Intergenerational ethics of high level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Kunihiko [Nagoya Univ., Graduate School of Engineering, Nagoya, Aichi (Japan); Nasu, Akiko; Maruyama, Yoshihiro [Shibaura Inst. of Tech., Tokyo (Japan)

    2003-03-01

    The validity of intergenerational ethics on the geological disposal of high level radioactive waste originating from nuclear power plants was studied. The result of the study on geological disposal technology showed that the current method of disposal can be judged to be scientifically reliable for several hundred years and the radioactivity level will be less than one tenth of the tolerable amount after 1,000 years or more. This implies that the consideration of intergenerational ethics of geological disposal is meaningless. Ethics developed in western society states that the consent of people in the future is necessary if the disposal has influence on them. Moreover, the ethics depends on generally accepted ideas in western society and preconceptions based on racism and sexism. The irrationality becomes clearer by comparing the dangers of the exhaustion of natural resources and pollution from harmful substances in a recycling society. (author)

  17. Intergenerational ethics of high level radioactive waste

    International Nuclear Information System (INIS)

    Takeda, Kunihiko; Nasu, Akiko; Maruyama, Yoshihiro

    2003-01-01

    The validity of intergenerational ethics on the geological disposal of high level radioactive waste originating from nuclear power plants was studied. The result of the study on geological disposal technology showed that the current method of disposal can be judged to be scientifically reliable for several hundred years and the radioactivity level will be less than one tenth of the tolerable amount after 1,000 years or more. This implies that the consideration of intergenerational ethics of geological disposal is meaningless. Ethics developed in western society states that the consent of people in the future is necessary if the disposal has influence on them. Moreover, the ethics depends on generally accepted ideas in western society and preconceptions based on racism and sexism. The irrationality becomes clearer by comparing the dangers of the exhaustion of natural resources and pollution from harmful substances in a recycling society. (author)

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

    International Nuclear Information System (INIS)

    Elmore, M.R.

    1996-02-01

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

  19. High polymer-based composites for the fabrication of containers for the long-term storage or disposal of high-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Miedema, I.; Bonin, H.W.; Bui, V.T. [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

    2002-07-01

    This study considers the application of PEEK and continuous graphite fibre composite as the principal component in a high level nuclear waste disposal container. The ultimate radioactive environment to which the containers will be exposed has been simulated using a SLOWPOKE-2 research nuclear reactor and a specialized heated irradiation chamber. Doses of up to 1 MGy were given to samples in combination with elevated temperatures (15{sup o}C to 75{sup o}C), which induced mechanical and chemical changes in the material. Mechanically, the composite and virgin polymer samples were minimally affected, rarely deviating beyond one standard deviation of the properties of unirradiated samples. Molecularly, crosslinking between adjacent polymer chains in the amorphous region is the primary observed phenomenon as a consequence of the radiation treatment. This effect is diminished with the application of heat during irradiation. Slight changes in crystallinity were also noted through molecular rearrangement, beginning with slight increases at lower radiation doses, and then minor decreases are noted with larger doses ({approx}10{sup 6} Gy). It is also shown in this study that the rate of radiation effects that is typical in this polymer is dependent on the temperature of irradiation. The results confirm that polymer-based composite materials, such as the PEEK/graphite fibre material studied here, are excellent candidates for the fabrication of the containers for the disposal of high-level radioactive waste. (author)

  20. DEFENSE HIGH LEVEL WASTE GLASS DEGRADATION

    International Nuclear Information System (INIS)

    Ebert, W.

    2001-01-01

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

  1. Evaluation of radionuclide concentrations in high-level radioactive wastes

    International Nuclear Information System (INIS)

    Fehringer, D.J.

    1985-10-01

    This report describes a possible approach for development of a numerical definition of the term ''high-level radioactive waste.'' Five wastes are identified which are recognized as being high-level wastes under current, non-numerical definitions. The constituents of these wastes are examined and the most hazardous component radionuclides are identified. This report suggests that other wastes with similar concentrations of these radionuclides could also be defined as high-level wastes. 15 refs., 9 figs., 4 tabs

  2. 40 CFR 227.30 - High-level radioactive waste.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 24 2010-07-01 2010-07-01 false High-level radioactive waste. 227.30...-level radioactive waste. High-level radioactive waste means the aqueous waste resulting from the operation of the first cycle solvent extraction system, or equivalent, and the concentrated waste from...

  3. Storage of long lived solid waste

    International Nuclear Information System (INIS)

    Ozarde, P.D.; Agarwal, K.; Gupta, R.K.; Gandhi, K.G.

    2009-01-01

    Long lived solid waste, generated during the fuel cycle mainly includes high level vitrified waste product, high level cladding hulls and low and intermediate level alpha wastes. These wastes require storage in specially designed engineered facilities before final disposal into deep geological repository. Since high-level vitrified waste contain heat generating radionuclides, the facility for their storage is designed for continuous cooling. High level cladding hulls undergo volume reduction by compaction and will be subsequently stored. (author)

  4. Project study for the final disposal of high-level radioactive wastes in deep-lying geological formations and for intermediate storage. Projektstudie fuer die Endlagerung von hochaktiven Abfaellen in tiefliegenden geologischen Formationen sowie fuer die Zwischenlagerung

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

    The present report has been written to show the feasibility of construction and the operational suitability of different intermediate and final storage concepts for high-level radioactive wastes. It summarizes the information from a project study given under contract by Nagra. The report should orient a broader public about the possible construction of such intermediate and final repositories. The work presented here refers to radioactive wastes which need an extremely long isolation time. Important information from a separate study concerning different aspects of the drilling of deep boreholes, has been integrated into the present report.

  5. Managing commercial high-level radioactive waste: summary

    International Nuclear Information System (INIS)

    1982-04-01

    This summary presents the findings and conclusions of OTA's analysis of Federal policy for the management of commercial high-level radioactive waste - an issue that has been debated over the last decade and that now appears to be moving toward major congressional action. After more than 20 years of commercial nuclear power, the Federal Government has yet to develop a broadly supported policy for fulfilling its legal responsibility for the final isolation of high-level radioactive waste. OTA's study concludes that until such a policy is adopted in law, there is a substantial risk that the false starts, shifts of policy, and fluctuating support that have plagued the final isolation program in the past will continue. The continued lack of final isolation facilities has raised two key problems that underlie debates about radioactive waste policy. First, some question the continued use of nuclear power until it is shown that safe final isolation for the resulting wastes can and will be accomplished, and argue that the failure to develop final isolation facilities is evidence that it may be an insoluble problem. Second, because there are no reprocessing facilities or federal waste isolation facilities to accept spent fuel, existing reactors are running out of spent fuel storage space, and by 1986 some may face a risk of shutting down for some period. Most of the 72,000 metric tons of spent fuel expected to be generated by the year 2000 will still be in temporary storage at that time. While it is possible that utilities could provide all necessary additional storage at reactor sites before existing basins are filled, some supplemental storage may be needed if there are delays in their efforts

  6. Managing commercial high-level radioactive waste

    International Nuclear Information System (INIS)

    1983-01-01

    The article is a summary of issues raised during US Congress deliberations on nuclear waste policy legislation. It is suggested that, if history is not to repeat itself, and the current stalemate on nuclear waste is not to continue, a comprehensive policy is needed that addresses the near-term problems of interim storage as part of an explicit and credible program for dealing with the longer term problem of developing a final isolation system. Such a policy must: 1) adequately address the concerns and win the support of all the major interested parties, and 2) adopt a conservative technical and institutional approach - one that places high priority on avoiding the problems that have repeatedly beset the program in the past. It is concluded that a broadly supported comprehensive policy would contain three major elements, each designed to address one of the key questions concerning Federal credibility: commitment in law to the goals of a comprehensive policy; credible institutional mechanisms for meeting goals; and credible measures for addressing the specific concerns of the states and the various publics. Such a policy is described in detail. (Auth.)

  7. High level waste at Hanford: Potential for waste loading maximization

    International Nuclear Information System (INIS)

    Hrma, P.R.; Bailey, A.W.

    1995-09-01

    The loading of Hanford nuclear waste in borosilicate glass is limited by phase-related phenomena, such as crystallization or formation of immiscible liquids, and by breakdown of the glass structure because of an excessive concentration of modifiers. The phase-related phenomena cause both processing and product quality problems. The deterioration of product durability determines the ultimate waste loading limit if all processing problems are resolved. Concrete examples and mass-balance based calculations show that a substantial potential exists for increasing waste loading of high-level wastes that contain a large fraction of refractory components

  8. National high-level waste systems analysis plan

    International Nuclear Information System (INIS)

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

    1995-05-01

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

  9. Decontamination of high-level waste canisters

    International Nuclear Information System (INIS)

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

    1980-12-01

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

  10. Material chemistry challenges in vitrification of high level radioactive waste

    International Nuclear Information System (INIS)

    Kaushik, C.P.

    2008-01-01

    Full text: Nuclear technology with an affective environmental management plan and focused attention on safety measures is a much cleaner source of electricity generation as compared to other sources. With this perspective, India has undertaken nuclear energy program to share substantial part of future need of power. Safe containment and isolation of nuclear waste from human environment is an indispensable part of this programme. Majority of radioactivity in the entire nuclear fuel cycle is high level radioactive liquid waste (HLW), which is getting generated during reprocessing of spent nuclear fuels. A three stage strategy for management of HLW has been adopted in India. This involves (i) immobilization of waste oxides in stable and inert solid matrices, (ii) interim retrievable storage of the conditioned waste product under continuous cooling and (iii) disposal in deep geological formation. Borosilicate glass matrix has been adopted in India for immobilization of HLW. Material issue are very important during the entire process of waste immobilization. Performance of the materials used in nuclear waste management determines its safety/hazards. Material chemistry therefore has a significant bearing on immobilization science and its technological development for management of HLW. The choice of suitable waste form to deploy for nuclear waste immobilization is difficult decision and the durability of the conditioned product is not the sole criterion. In any immobilization process, where radioactive materials are involved, the process and operational conditions play an important role in final selection of a suitable glass formulation. In remotely operated vitrification process, study of chemistry of materials like glass, melter, materials of construction of other equipment under high temperature and hostile corrosive condition assume significance for safe and un-interrupted vitrification of radioactive to ensure its isolation waste from human environment. The present

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

    International Nuclear Information System (INIS)

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

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

  12. On risk assessment of high level radioactive waste disposal

    International Nuclear Information System (INIS)

    Smith, C.F.; Kastenberg, W.E.

    1976-01-01

    One of the major concerns with the continued growth of the nuclear power industry is the production of the high level radioactive wastes. The risks associated with the disposal of these wastes derives from the potential for release of radioactive materials into the environment. The development of a methodology for risk analysis is carried out. The methodology suggested involves the probabilistic analysis of a general accident consequence distribution. In this analysis, the frequency aspect of the distribution is treated separately from the normalized probability function. In the final stage of the analysis, the frequency and probability characteristics of the distribution are recombined to provide an estimate of the risk. The characterization of the radioactive source term is accomplished using the ORIGEN computer code. Calculations are carried out for various reactor types and fuel cycles, and the overall waste hazard for a projected 35 year nuclear power program is determined. An index of relative nuclide hazard appropriate to problems involving the management of high level radioactive wastes is developed. As an illustration of the methodology, risk analyses are made for two proposed methods for waste management: extraterrestrial disposal and interim surface storage. The results of these analyses indicate that, within the assumptions used, the risks of these management schemes are small compared with natural background radiation doses. (Auth.)

  13. Conceptual process for conversion of high level waste to glass

    International Nuclear Information System (INIS)

    1975-01-01

    During a ten-year period highly radioactive wastes amounting to 22 million gallons of salt cake and 5 million gallons of wet sludge are to be converted to 1.2 million gallons of glass and 24 million gallons of decontaminated salt cake and placed in the new storage facilities which will provide high assurance of containment with minimal reliance on maintenance and surveillance. The glass will contain nearly all of the radioactivity in a form that is highly resistant to leaching and dispersion. The salt cake will contain a small amount of residual radioactivity. The process is shown in Figure 1 and the facilities may be arranged in seven modules to accomplish seven tasks, (1) remove wastes from tanks, (2) separate sludge and salt, (3) decontaminate salt, (4) solidify and package sludge and 137 Cs, (5) solidify and package decontaminated salt, (6) store high level waste, and (7) store decontaminated salt cake

  14. Technologies for recovery of transuranics and immobilization of non-high-level wastes

    International Nuclear Information System (INIS)

    Richardson, G.L.

    1976-06-01

    This paper supplements the preceding Symposium paper on ''Treatment Technologies for Non-High-Level Wastes (U.S.A.)'' by C. R. Cooley and D. E. Clark (HEDL-SA-851), and covers the additional treatment technologies in use and under development for recovering transuranics and immobilizing non-high-level wastes for transportation and storage. Methods used for nondestructive assay (NDA) of TRU elements in non-high-level wastes are also discussed briefly

  15. Application of SYNROC to high-level defense wastes

    International Nuclear Information System (INIS)

    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 phase 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 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 FY81

  16. High-level waste melter alternatives assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Calmus, R.B.

    1995-02-01

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

  17. High-level waste melter alternatives assessment report

    International Nuclear Information System (INIS)

    Calmus, R.B.

    1995-02-01

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

  18. The high level and long lived radioactive wastes

    International Nuclear Information System (INIS)

    2005-01-01

    This report presents the main conclusions of 15 years of researches managed by the CEA. This report is the preliminary version of the 2005 final report. It presents the main conclusions of the actions on the axis 1 and 3 of the law of the 30 December 1991. The synthesis report on the axis 1 concerns results obtained on the long lived radionuclides separation and transmutation in high level and long lived radioactive wastes. the synthesis report on the axis 3 presents results obtained by the processes of conditioning and of ground and underground long term storage. (A.L.B.)

  19. Stability of High-Level Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Solidification of high-level radioactive wastes. Final report

    International Nuclear Information System (INIS)

    1979-06-01

    A panel on waste solidification was formed at the request of the Nuclear Regulatory Commission to study the scientific and technological problems associated with the conversion of liquid and semiliquid high-level radioactive wastes into a stable form suitable for transportation and disposition. Conclusions reached and recommendations made are as follows. Many solid forms described in this report could meet standards as stringent as those currently applied to the handling, storage, and transportation of spent fuel assemblies. Solid waste forms should be selected only in the context of the total radioactive waste management system. Many solid forms are likely to be satisfactory for use in an appropriately designed system, The current United States policy of deferring the reprocessing of commercial reactor fuel provides additional time for R and D solidification technology for this class of wastes. Defense wastes which are relatively low in radioactivity and thermal power density can best be solidified by low-temperature processes. For solidification of fresh commercial wastes that are high in specific activity and thermal power density, the Panel recommends that, in addition to glass, the use of fully-crystalline ceramics and metal-matrix forms be actively considered. Preliminary analysis of the characteristics of spent fuel pins indicates that they may be eligible for consideration as a waste form. Because the differences in potential health hazards to the public resulting from the use of various solid form and disposal options are likely to be small, the Panel concludes that cost, reliability, and health hazards to operating personnel will be major considerations in choosing among the options that can meet safety requiremens. The Panel recommends that responsibility for all radioactive waste management operations (including solidification R and D) should be centralized

  1. High-level waste tank farm set point document

    International Nuclear Information System (INIS)

    Anthony, J.A. III.

    1995-01-01

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

  2. High-level waste tank farm set point document

    Energy Technology Data Exchange (ETDEWEB)

    Anthony, J.A. III

    1995-01-15

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

  3. Managing the nation's commercial high-level radioactive waste

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    This study presents the findings and conclusions of OTA's analysis of Federal policy for the management of commercial high-level radioactive waste. Broad in scope and balanced in approach, its coverage extends from technological and organizational questions to political ramifications...the environmental impact of building repositories...and even dealing with Indian tribes affected by repository site selection and development. Emphasis is on workable strategies for implementing the National Waste Policy Act of 1982, including a mission plan for the program...a monitored retrievable storage proposal...and a report on mechanisms for financing and managing the program. Nine appendicies are included. They furnish additional data on such topics as policymaking, history, and the system issues resolved in NWPA

  4. Standards for high level waste disposal: A sustainability perspective

    International Nuclear Information System (INIS)

    Dougherty, W.W.; Powers, V.; Johnson, F.X.; Cornland, D.

    1999-01-01

    Spent reactor fuel from commercial power stations contains high levels of plutonium, other fissionable actinides, and fission products, all of which pose serious challenges for permanent disposal because of the very long half-lives of some isotopes. The 'nuclear nations' have agreed on the use of permanent geologic repositories for the ultimate disposal of high-level nuclear waste. However, it is premature to claim that a geologic repository offers permanent isolation from the biosphere, given high levels of uncertainty, nascent risk assessment frameworks for the time periods considered, and serious intergenerational equity issues. Many have argued for a broader consideration of disposal options that include extended monitored retrievable storage and accelerator-driven transmutation of wastes. In this paper we discuss and compare these three options relative to standards that emerge from the application of sustainable development principles, namely long-lasting technical viability, intergenerational equity, rational resource allocation, and rights of future intervention. We conclude that in order to maximise the autonomy of future generations, it is imperative to leave future options more open than does permanent disposal

  5. Disposal of high-level radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Costello, J M [Australian Atomic Energy Commission Research Establishment, Lucas Heights

    1982-03-01

    The aims and options for the management and disposal of highly radioactive wastes contained in spent fuel from the generation of nuclear power are outlined. The status of developments in reprocessing, waste solidification and geologic burial in major countries is reviewed. Some generic assessments of the potential radiological impacts from geologic repositories are discussed, and a perspective is suggested on risks from radiation.

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

    International Nuclear Information System (INIS)

    Collins, K.J.

    1998-01-01

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

  7. Salt removal from tanks containing high-level radioactive waste

    International Nuclear Information System (INIS)

    Kiser, D.L.

    1981-01-01

    At the Savannah River Plant (SRP), there are 23 waste storage tanks containing high-level radioactive wastes that are to be retired. These tanks contain about 23 million liters of salt and about 10 million liters of sludge, that are to be relocated to new Type III, fully stress-relieved tanks with complete secondary containment. About 19 million liters of salt cake are to be dissolved. Steam jet circulators were originally proposed for the salt dissolution program. However, use of steam jet circulators raised the temperature of the tank contents and caused operating problems. These included increased corrosion risk and required long cooldown periods prior to transfer. Alternative dissolution concepts were investigated. Examination of mechanisms affecting salt dissolution showed that the ability of fresh water to contact the cake surface was the most significant factor influencing dissolution rate. Density driven and mechanical agitation techniques were developed on a bench scale and then were demonstrated in an actual waste tank. Actual waste tank demonstrations were in good agreement with bench-scale experiments at 1/85 scale. The density driven method utilizes simple equipment, but leaves a cake heel in the tank and is hindered by the presence of sludge or Zeolite in the salt cake. Mechanical agitation overcomes the problems found with both steam jet circulators and the density driven technique and is the best method for future waste tank salt removal

  8. High level waste management in France

    International Nuclear Information System (INIS)

    Sombret, C.; Bonniaud, R.; Jouan, A.

    1979-01-01

    This paper deals with solidification of radioactive wastes. The vitrification process has been selected, the Marcoule vitrification plant and the operation are described. Prospects for vitrification of fission product solutions from La Hague reprocessing plant are given

  9. Idaho National Engineering Laboratory High-Level Waste Roadmap

    International Nuclear Information System (INIS)

    1993-08-01

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

  10. High-level waste immobilization program: an overview

    International Nuclear Information System (INIS)

    Bonner, W.R.

    1979-09-01

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

  11. Multipurpose optimization models for high level waste vitrification

    International Nuclear Information System (INIS)

    Hoza, M.

    1994-08-01

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

  12. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Science.gov (United States)

    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 C...

  13. Licensing information needs for a high-level waste repository

    International Nuclear Information System (INIS)

    Wright, R.J.; Greeves, J.T.; Logsdon, M.J.

    1985-01-01

    The information needs for licensing findings during the development of a repository for high-level waste (HLW) are described. In particular, attention is given to the information and needs to demonstrate, for construction authorization purposes: repository constructibility, waste retrievability, waste containment, and waste isolation

  14. High-level waste-form-product performance evaluation

    International Nuclear Information System (INIS)

    Bernadzikowski, T.A.; Allender, J.S.; Stone, J.A.; Gordon, D.E.; Gould, T.H. Jr.; Westberry, C.F. III.

    1982-01-01

    Seven candidate waste forms were evaluated for immobilization and geologic disposal of high-level radioactive wastes. The waste forms were compared on the basis of leach resistance, mechanical stability, and waste loading. All forms performed well at leaching temperatures of 40, 90, and 150 0 C. Ceramic forms ranked highest, followed by glasses, a metal matrix form, and concrete. 11 tables

  15. Long-term high-level waste technology program

    International Nuclear Information System (INIS)

    1980-04-01

    The Department of Energy (DOE) is conducting a comprehensive program to isolate all US nuclear wastes from the human environment. The DOE Office of Nuclear Energy - Waste (NEW) has full responsibility for managing the high-level wastes resulting from defense activities and additional responsiblity for providing the technology to manage existing commercial high-level wastes and any that may be generated in one of several alternative fuel cycles. Responsibilities of the Three Divisions of DOE-NEW are shown. This strategy document presents the research and development plan of the Division of Waste Products for long-term immobilization of the high-level radioactive wastes resulting from chemical processing of nuclear reactor fuels and targets. These high-level wastes contain more than 99% of the residual radionuclides produced in the fuels and targets during reactor operations. They include essentially all the fission products and most of the actinides that were not recovered for use

  16. Performance of evaporators in high level radioactive chemical waste service

    International Nuclear Information System (INIS)

    Jenkins, C.F.

    1997-01-01

    Chemical processing of nuclear fuels and targets at Savannah River Site resulted in generation of millions of gallons of liquid wastes. The wastes were further processed to reduce volume and allow for extended temporary storage of a more concentrated material. Waste evaporators have been a central point for waste reduction for many years. Currently, the transfer and processing of the concentrated wastes for permanent storage requires dilution and results in generation of significant quantities of additional liquid wastes. A new round of volume reduction is required to fit existing storage capacity and to allow for removal of older vessels from service. Evaporator design, performance and repairs are discussed in this report

  17. Characteristics of solidified high-level waste products

    International Nuclear Information System (INIS)

    1979-01-01

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

  18. Safety relevant aspects of the long-term intermediate storage of spent fuel elements and vitrified high-level radioactive wastes; Sicherheitstechnische Aspekte der langfristigen Zwischenlagerung von bestrahlten Brennelementen und verglastem HAW

    Energy Technology Data Exchange (ETDEWEB)

    Ellinger, A.; Geupel, S.; Gewehr, K.; Gmal, B.; Hannstein, V.; Hummelsheim, K.; Kilger, R.; Wagner, M. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany); Schmidt, G.; Spieth-Achtnich, A. [Oeko-Institut e.V. - Institut fuer angewandte Oekolgie (Germany)

    2010-04-15

    The currently in Germany pursued concept for management of spent fuel from nuclear power plants provides intermediate dry cask storage at the NPP sites until direct disposal in a deep geologic repository. In addition the earlier commissioned centralized dry storage facilities are being used for storage of high level radioactive waste returned from foreign reprocessing of German spent fuel performed so far. The dry interim storage facilities are licensed for 40 years of operation time. According to the German regulations a full scope periodic safety review is not required so far, neither practical experience on dry storage for this period of time is available. With regard to this background the report at hand is dealing with long term effects, which may affect safety of the interim storage during the 40 years period or beyond if appropriate, and with the question, whether additional analyses or monitoring measures may be required. Therefore relevant publications have been evaluated, calculations have been performed as well as a systematic screening with regard to loads and possible ageing effects has been applied to structures and components important for safety of the storage, in order to identify relevant long term effects, which may not have been considered sufficiently so far and to provide proposals for an improved ageing management. The report firstly provides an overview on the current state of technology describing shortly the national and international practice and experience. In the following chapters safety aspects of interim storage with regard to time dependent effects and variations are being analyzed and discussed. Among this not only technical aspects like the long term behavior of fuel elements, canisters and storage systems are addressed, but also operational long term aspects regarding personnel planning, know how conservation, documentation and quality management are taken into account. A separate chapter is dedicated to developing and describing

  19. Radiolytic gas formation in high-level liquid waste solutions

    International Nuclear Information System (INIS)

    Brodda, B.-G.; Dix, Siegfried; Merz, E.R.

    1989-01-01

    High-level fission product waste solutions originating from the first-cycle raffinate stream of spent fast breeder reactor fuel reprocessing have been investigated gas chromatographically for their radiolytic and chemical gas production. The solutions showed considerable formation of hydrogen, carbon dioxide and dinitrogen oxide, whereas atmospheric oxygen was consumed completely within a short time. In particular, carbon dioxide resulted from the radiolytic degradation of entrained organic solvent. After nearly complete degradation of the organic solvent, the influence of hydrazine and nitrogen dioxide on hydrogen formation was investigated. Hydrazinium hydroxide led to the formation of dinitrogen oxide and nitrogen. After 60 d, the concentration of dinitrogen oxide had reduced to zero, whereas the amount of nitrogen formed had reached a maximum. This may be explained by simultaneous chemical and radiolytic reactions leading to the formation of dinitrogen oxide and nitrogen and photolytic fission of dinitrogen oxide. Addition of sodium nitrite resulted in the rapid formation of dinitrogen oxide. The rate of hydrogen production was not changed significantly after the addition of hydrazine or nitrite. The results indicate that under normal operating conditions no dangerous hydrogen radiolysis yields should develop in the course of reprocessing and high-level liquid waste tank storage. Organic entrainment may lead to enhanced radiolytic decomposition and thus to considerable hydrogen production rates and pressure build-up in closed systems. (author)

  20. Radioactive waste interim storage in Germany

    International Nuclear Information System (INIS)

    2015-12-01

    The short summary on the radioactive waste interim storage in Germany covers the following issues: importance of interim storage in the frame of radioactive waste management, responsibilities and regulations, waste forms, storage containers, transport of vitrified high-level radioactive wastes from the reprocessing plants, central interim storage facilities (Gorleben, Ahaus, Nord/Lubmin), local interim storage facilities at nuclear power plant sites, federal state collecting facilities, safety, radiation exposure in Germany.

  1. Sustainable Development and High Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Mikael [Swedish Radiation Protection Inst., Stockholm (Sweden)

    2001-07-01

    Sustainable development, defined by the BrundtIand Commission as 'development that meets the needs of the present without compromising the ability of future generations to meet their own needs', relates to a number of issues such as population, health, food, species and ecosystems, energy, industrial development, urbanization, societal issues and economy, and how these global challenges could be met within a long term strategy. It is not obvious how the principle may be applied to final disposal of radioactive waste, but the global scope of the principle suggests that no sector in society should be exempted from scrutinizing its practices in the light of the challenge presented by sustainable development. Waste management, as pointed out by the International Commission on Radiological Protection, ICRP, cannot be seen as a free standing practice in need of its own justification. The produced waste cannot be seen separately from the other components of nuclear production. However, the existence of very long-lived radioactive nuclei in the spent fuel warrants a careful examination of this subpractice. Health based post-closure criteria or standards for long-lived waste, usually make use of the concept of partitioning dose limit. ICRP recommends that individuals in the public do not receive a yearly dose in excess of 1 mSv as a result of releases in connection with activities involving the use of ionising radiation, and that any single facility does not generate a dose burden to individuals in excess of a fraction of this value. For an operating facility, this fraction is normally at least a factor of three. By definition, operational changes are not possible for a closed repository. It follows from this that the partitioning has another function. One interpretation is that it can allow for the simultaneous use and burdens of future generation's activities. Both the Swedish and the proposed US criteria and from EPA and NRC, as well as standards from

  2. Sustainable Development and High Level Waste

    International Nuclear Information System (INIS)

    Jensen, Mikael

    2001-01-01

    Sustainable development, defined by the BrundtIand Commission as 'development that meets the needs of the present without compromising the ability of future generations to meet their own needs', relates to a number of issues such as population, health, food, species and ecosystems, energy, industrial development, urbanization, societal issues and economy, and how these global challenges could be met within a long term strategy. It is not obvious how the principle may be applied to final disposal of radioactive waste, but the global scope of the principle suggests that no sector in society should be exempted from scrutinizing its practices in the light of the challenge presented by sustainable development. Waste management, as pointed out by the International Commission on Radiological Protection, ICRP, cannot be seen as a free standing practice in need of its own justification. The produced waste cannot be seen separately from the other components of nuclear production. However, the existence of very long-lived radioactive nuclei in the spent fuel warrants a careful examination of this subpractice. Health based post-closure criteria or standards for long-lived waste, usually make use of the concept of partitioning dose limit. ICRP recommends that individuals in the public do not receive a yearly dose in excess of 1 mSv as a result of releases in connection with activities involving the use of ionising radiation, and that any single facility does not generate a dose burden to individuals in excess of a fraction of this value. For an operating facility, this fraction is normally at least a factor of three. By definition, operational changes are not possible for a closed repository. It follows from this that the partitioning has another function. One interpretation is that it can allow for the simultaneous use and burdens of future generation's activities. Both the Swedish and the proposed US criteria and from EPA and NRC, as well as standards from Canada, UK and

  3. Options for the disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    Mitchell, N.T.; Laughton, A.S.; Webb, G.A.M.

    1977-01-01

    The management of radioactive waste within the fuel cycle, especially the high-level wastes from reprocessing of nuclear fuel, is currently a matter of particular concern. In the short term (meaning a timescale of tens of years) management by engineered storage is considered to provide a satisfactory solution. Beyond this, however, the two main alternative options which are considered in the paper are: (a) disposal by burial into geologic formations on land; and (b) disposal by emplacement into or onto the seabed. Status of our present knowledge on the land and seabed disposal options is reviewed together with an assessment of the extent to which their reliability and safety can be judged on presently available information. Further information is needed on the environmental behaviour of radioactivity in the form of solidified waste in both situations in order to provide a more complete, scientific assessment. Work done so far has clarified the areas where further research is most needed - for instance modelling of the environmental transfer processes associated with the seabed option. This is discussed together with an indication of the research programmes which are now being pursued

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  5. High-level radioactive waste fixation in sintered vitreous matrix

    International Nuclear Information System (INIS)

    Russo, D.O.; Messi de Bernasconi, N.; Audero, M.A.

    1987-01-01

    The safe storage of high-level wastes from fuel elements reprocessing includes, as a first step, the fixation of the same in materials having a good resistance to the leaching in aqueous medium, such as borosilicate glass. As an alternative to the usual method of the molten glasses, a procedure for the sintering of a powdered glass and waste mixture at lower temperatures (600-700 deg C) has been developed, which minimizes the volatilization of active compounds during the process. Two glasses matrices of different composition and characteristics were used, to which the simulated wastes were added in the ratio of a 10% in weight of oxides. Two sintering techniques were employed 1: cold pressing and further sintering; 2: hot pressing and sintering under pressure. The densities were measured, the microstructure of the samples was analyzed and leaching essays were made in distilled water. The pellet's microstructure was observed by means of optical microscopy, by reflection in polished samples and by transparency in thin slices. The presence of crystalline compounds was analyzed by means of x rays and electron microprobe. The results have shown the convenience to continue with hot pressing essays, because a denser product with a higher resistance to the leaching is thus obtained. (M.E.L.) [es

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

    International Nuclear Information System (INIS)

    Stegen, G.E.

    1996-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  8. Neutron and gamma-ray sources in LWR high-level nuclear waste

    International Nuclear Information System (INIS)

    Dupree, S.A.

    1977-06-01

    Predictions of the composition of high-level waste from U-fueled LWRs have been used to calculate the neutron and gamma-ray sources in such waste at cooling times of 3 and 10 years. The results are intended for interim application to studies of waste shipping and storage pending the availability of more exact knowledge of fuel recycling and of waste concentration and solidification

  9. Handbook of high-level radioactive waste transportation

    International Nuclear Information System (INIS)

    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

  10. The immobilization of High Level Waste Into Glass

    International Nuclear Information System (INIS)

    Aisyah; Martono, H.

    1998-01-01

    High level liquid waste is generated from the first step extraction in the nuclear fuel reprocessing. The waste is immobilized with boro-silicate glass. A certain composition of glass is needed for a certain type of waste, so that the properties of waste glass would meet the requirement either for further process or for disposal. The effect of waste loading on either density, thermal expansion, softening point and leaching rate has been studied. The composition of the high level liquid waste has been determined by ORIGEN 2 and the result has been used to prepare simulated high level waste. The waste loading in the waste glass has been set to be 19.48; 22.32; 25.27; and 26.59 weight percent. The result shows that increasing the waste loading has resulted in the higher density with no thermal expansion and softening point significant change. The increase in the waste loading increase that leaching rate. The properties of the waste glass in this research have not shown any deviation from the standard waste glass properties

  11. Techniques for the solidification of high-level wastes

    International Nuclear Information System (INIS)

    1977-01-01

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

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

    International Nuclear Information System (INIS)

    1982-03-01

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

  13. ONDRAF/NIRAS and high-level radioactive waste management in Belgium

    International Nuclear Information System (INIS)

    Decamps, F.

    1993-01-01

    The National Agency for Radioactive Waste and Enriched Fissile Materials, ONDRAF/NIRAS, is a public body with legal personality in charge of managing all radioactive waste on Belgian territory, regardless of its origin and source. It is also entrusted with tasks related to the management of enriched fissile materials, plutonium containing materials and used or unused nuclear fuel, and with certain aspects of the dismantling of closed down nuclear facilities. High-level radioactive waste management comprises essentially and for the time being the storage of high-level liquid waste produced by the former EUROCHEMIC reprocessing plant and of high-level and very high-level heat producing waste resulting from the reprocessing in France of Belgian spent fuel, as well as research and development (R and D) with regard to geological disposal in clay of this waste type

  14. Final report on cermet high-level waste forms

    International Nuclear Information System (INIS)

    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

  15. Evaluation of solidified high-level waste forms

    International Nuclear Information System (INIS)

    1981-01-01

    One of the objectives of the IAEA waste management programme is to coordinate and promote development of improved technology for the safe management of radioactive wastes. The Agency accomplished this objective specifically through sponsoring Coordinated Research Programmes on the ''Evaluation of Solidified High Level Waste Products'' in 1977. The primary objectives of this programme are to review and disseminate information on the properties of solidified high-level waste forms, to provide a mechanism for analysis and comparison of results from different institutes, and to help coordinate future plans and actions. This report is a summary compilation of the key information disseminated at the second meeting of this programme

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

    Science.gov (United States)

    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

  17. Technical career opportunities in high-level radioactive waste management

    International Nuclear Information System (INIS)

    1993-01-01

    Technical career opportunities in high-level radioactive waste management are briefly described in the areas of: Hydrology; geology; biological sciences; mathematics; engineering; heavy equipment operation; and skilled labor and crafts

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

    International Nuclear Information System (INIS)

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

    1997-09-01

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

  19. High-Level Waste System Process Interface Description

    International Nuclear Information System (INIS)

    D'Entremont, P.D.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Cornman, W.R.

    1981-08-01

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

  1. DOUBLE SHELL TANK INTEGRITY PROJECT HIGH LEVEL WASTE CHEMISTRY OPTIMIZATION

    International Nuclear Information System (INIS)

    WASHENFELDER DJ

    2008-01-01

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

  2. High-level radioactive waste disposal problem in Russia

    International Nuclear Information System (INIS)

    Velichkin, Vasily I.

    1999-01-01

    This presentation on radioactive waste management in Russia discusses criteria for the selection of disposal sites, how the various types of waste should be contained and stored, and gives a list showing the liable owner, type, volume, activity and storage place of the present amount of radioactive waste. The bulk of this waste, in volume and radioactivity, is at the enterprises of Minatom of the Russian Federation

  3. Radiation transport in high-level waste form

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    D.C. Richardson

    2003-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    Kudo, Koji; Takada, Susumu; Kawanishi, Motoi

    2005-01-01

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

  7. High-level radioactive waste disposal type and theoretical analyses

    International Nuclear Information System (INIS)

    Lu Yingfa; Wu Yanchun; Luo Xianqi; Cui Yujun

    2006-01-01

    Study of high-level radioactive waste disposal is necessary for the nuclear electrical development; the determination of nuclear waste depository type is one of importance safety. Based on the high-level radioactive disposal type, the relative research subjects are proposed, then the fundamental research characteristics of nuclear waste disposition, for instance: mechanical and hydraulic properties of rock mass, saturated and unsaturated seepage, chemical behaviors, behavior of special soil, and gas behavior, etc. are introduced, the relative coupling equations are suggested, and a one dimensional result is proposed. (authors)

  8. Cermet high level waste forms: a pregress report

    International Nuclear Information System (INIS)

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

    1978-06-01

    The fixation of high level radioactive waste from both commercial and DOE defense sources as cermets is currently under study. This waste form consists of a continuous iron-nickel base metal matrix containing small particles of fission product oxides. Preliminary evaluations of cermets fabricated from a variety of simulated wastes indicate they possess properties providing advantages over other waste forms presently being considered, namely thermal conductivity, waste loading levels, and leach resistance. This report describes the progress of this effort, to date, since its initiation in 1977

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

    International Nuclear Information System (INIS)

    Cornman, W.R.

    1981-02-01

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

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

    International Nuclear Information System (INIS)

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

    1994-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-05-01

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

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

    International Nuclear Information System (INIS)

    Crawford, T.W.

    1995-01-01

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

  13. Managing the nation's commercial high-level radioactive waste

    International Nuclear Information System (INIS)

    1985-03-01

    This report presents the findings and conclusions of OTA's analysis of Federal policy for the management of commercial high-level radioactive waste. It is intended to contribute to the implementation of Nuclear Waste Policy Act of 1982 (NWPA). The major conclusion of that review is that NWPA provides sufficient authority for developing and operating a waste management system based on disposal in geologic repositories. Substantial new authority for other facilities will not be required unless major unexpected problems with geologic disposal are encountered. OTA also concludes that DOE's Draft Mission Plan published in 1984 falls short of its potential for enhancing the credibility and acceptability of the waste management program

  14. Solidification of Savannah River Plant high-level waste

    International Nuclear Information System (INIS)

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

    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

  15. Production and properties of solidified high-level waste

    International Nuclear Information System (INIS)

    Brodersen, K.

    1980-08-01

    Available information on production and properties of solidified high-level waste are presented. The review includes literature up to the end of 1979. The feasibility of production of various types of solidified high-level wast is investigated. The main emphasis is on borosilicate glass but other options are also mentioned. The expected long-term behaviour of the materials are discussed on the basis of available results from laboratory experiments. Examples of the use of the information in safety analysis of disposal in salt formations are given. The work has been made on behalf of the Danish utilities investigation of the possibilities of disposal of high-level waste in salt domes in Jutland. (author)

  16. Evaluation of conditioned high-level waste forms

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  17. Defense High Level Waste Disposal Container System Description Document

    International Nuclear Information System (INIS)

    Pettit, N. E.

    2001-01-01

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

  18. High level radioactive waste management facility design criteria

    International Nuclear Information System (INIS)

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

    1993-01-01

    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

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

    International Nuclear Information System (INIS)

    STAEHR, T.W.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

    This patent describes lead-iron phosphate glasses containing a high level of Fe 2 O 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 0 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 2 to 10 3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe 2 O 3 in forming the lead-iron phosphate glass is critical. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms

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

    International Nuclear Information System (INIS)

    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

  2. In-situ nitrite analysis in high level waste tanks

    International Nuclear Information System (INIS)

    O'Rourke, P.E.; Prather, W.S.; Livingston, R.R.

    1992-01-01

    The Savannah River Site produces special nuclear materials used in the defense of the United States. Most of the processes at SRS are primarily chemical separations and purifications. In-situ chemical analyses help improve the safety, efficiency and quality of these operations. One area where in situ fiberoptic spectroscopy can have a great impact is the management of high level radioactive waste. High level radioactive waste at SRS is stored in more than 50 large waste tanks. The waste exists as a slurry of nitrate salts and metal hydroxides at pH's higher than 10. Sodium Nitrite is added to the tanks as a corrosion inhibitor. In-situ fiberoptic probes are being developed to measure the nitrate, nitrite and hydroxide concentrations in both liquid and solid fractions. Nitrite levels can be measured between 0.01M and 1M in a 1mm pathlength optical cell

  3. Development of high-level waste solidification technology 1

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Joon Hyung; Kim, Hwan Young; Kim, In Tae [and others

    1999-02-01

    Spent nuclear fuel contains useful nuclides as valuable resource materials for energy, heat and catalyst. High-level wastes (HLW) are expected to be generated from the R and D activities and reuse processes. It is necessary to develop vitrification or advanced solidification technologies for the safe long-term management of high level wastes. As a first step to establish HLW vitrification technology, characterization of HLWs that would arise at KAERI site, glass melting experiments with a lab-scale high frequency induction melter, and fabrication and property evaluation of base-glass made of used HEPA filter media and additives were performed. Basic study on the fabrication and characterization of candidate ceramic waste form (Synroc) was also carried out. These HLW solidification technologies would be directly useful for carrying out the R and Ds on the nuclear fuel cycle and waste management. (author). 70 refs., 29 tabs., 35 figs.

  4. Interim storage of radioactive waste packages

    International Nuclear Information System (INIS)

    1998-01-01

    This report covers all the principal aspects of production and interim storage of radioactive waste packages. The latest design solutions of waste storage facilities and the operational experiences of developed countries are described and evaluated in order to assist developing Member States in decision making and design and construction of their own storage facilities. This report is applicable to any category of radioactive waste package prepared for interim storage, including conditioned spent fuel, high level waste and sealed radiation sources. This report addresses the following issues: safety principles and requirements for storage of waste packages; treatment and conditioning methods for the main categories of radioactive waste; examples of existing interim storage facilities for LILW, spent fuel and high level waste; operational experience of Member States in waste storage operations including control of storage conditions, surveillance of waste packages and observation of the behaviour of waste packages during storage; retrieval of waste packages from storage facilities; technical and administrative measures that will ensure optimal performance of waste packages subject to various periods of interim storage

  5. Engineering-scale vitrification of commercial high-level waste

    International Nuclear Information System (INIS)

    Bonner, W.F.; Bjorklund, W.J.; Hanson, M.S.; Knowlton, D.E.

    1980-04-01

    To date, technology for immobilizing commercial high-level waste (HLW) has been extensively developed, and two major demonstration projects have been completed, the Waste Solidification Engineering Prototypes (WSEP) Program and the Nuclear Waste Vitrification Project (NWVP). The feasibility of radioactive waste solidification was demonstrated in the WSEP program between 1966 and 1970 (McElroy et al. 1972) using simulated power-reactor waste composed of nonradioactive chemicals and HLW from spent, Hanford reactor fuel. Thirty-three engineering-scale canisters of solidified HLW were produced during the operations. In early 79, the NWVP demonstrated the vitrification of HLW from the processing of actual commercial nuclear fuel. This program consisted of two parts, (1) waste preparation and (2) vitrification by spray calcination and in-can melting. This report presents results from the NWVP

  6. Final disposal of high levels waste and spent nuclear fuel

    International Nuclear Information System (INIS)

    Gelin, R.

    1984-05-01

    Foreign and international activities on the final disposal of high-level waste and spent nuclear fuel have been reviewed. A considerable research effort is devoted to development of acceptable disposal options. The different technical concepts presently under study are described in the report. Numerous studies have been made in many countries of the potential risks to future generations from radioactive wastes in underground disposal repositories. In the report the safety assessment studies and existing performance criteria for geological disposal are briefly discussed. The studies that are being made in Canada, the United States, France and Switzerland are the most interesting for Sweden as these countries also are considering disposal into crystalline rocks. The overall time-tables in different countries for realisation of the final disposal are rather similar. Normally actual large-scale disposal operations for high-level wastes are not foreseen until after year 2000. In the United States the Congress recently passed the important Nuclear Waste Policy Act. It gives a rather firm timetable for site-selection and construction of nuclear waste disposal facilities. According to this act the first repository for disposal of commercial high-level waste must be in operation not later than in January 1998. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Review of high-level waste form properties

    International Nuclear Information System (INIS)

    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

  9. High level waste management in Asia: R and D perspectives

    International Nuclear Information System (INIS)

    Deokattey, Sangeeta; Bhanumurthy, K.

    2010-01-01

    The present work is an attempt to provide an overview, about the status of R and D and current trends in high level radioactive waste management, particularly in Asian countries. The INIS database (for the period 1976 to 2010) was selected for this purpose, as this is the most authoritative global source of information, in the area of Nuclear Science and Technology. Appropriate query formulations on the database, resulted in the retrieval of 4322 unique bibliographic records. Using the content analysis method (which is both a qualitative as well as a quantitative research method), all the records were analyzed. Part One of the analysis details Scientometric R and D indicators, such as the countries and the institutions involved in R and D, the types of publications, and programmes and projects related to High Level Waste management. Part Two is a subject-based analysis, grouped under the following broad categories: I. Waste Processing 1. Partitioning and transmutation (including ADS) II. Waste Immobilization 1. Glass waste forms and 2. Crystalline ceramics and other waste forms III. Waste Disposal 1. Performance assessment and safety evaluation studies 2. Geohydrological studies a. Site selection and characterization, b. In situ underground experiments, c. Rock mechanical characterization 3. Deep geological repositories a. Sorption, migration and groundwater chemistry b. Engineered barrier systems and IV. Waste Packaging Materials. The results of this analysis are summarized in the study. (author)

  10. Disposal of high level and intermediate level radioactive wastes

    International Nuclear Information System (INIS)

    Flowers, R.H.

    1991-01-01

    The waste products from the nuclear industry are relatively small in volume. Apart from a few minor gaseous and liquid waste streams, containing readily dispersible elements of low radiotoxicity, all these products are processed into stable solid packages for disposal in underground repositories. Because the volumes are small, and because radioactive wastes are latecomers on the industrial scene, a whole new industry with a world-wide technological infrastructure has grown up alongside the nuclear power industry to carry out the waste processing and disposal to very high standards. Some of the technical approaches used, and the Regulatory controls which have been developed, will undoubtedly find application in the future to the management of non-radioactive toxic wastes. The repository site outlined would contain even high-level radioactive wastes and spent fuels being contained without significant radiation dose rates to the public. Water pathway dose rates are likely to be lowest for vitrified high-level wastes with spent PWR fuel and intermediate level wastes being somewhat higher. (author)

  11. High-level radioactive waste management in the United States. Background and status: 1996

    International Nuclear Information System (INIS)

    Dyer, J.R.

    1996-01-01

    The US high-level radioactive waste disposal program is investigating a site at Yucca Mountain, Nevada, to determine whether or not it is a suitable location for the development of a deep mined geologic repository. At this time, the US program is investigating a single site, although in the past, the program involved successive screening and comparison of alternate locations. The United States civilian reactor programs do not reprocess spent fuel; the high-level waste repository will be designed for the emplacement or spent fuel and a limited amount of vitrified high-level wastes from previous reprocessing in the US. The legislation enabling the US program also contains provisions for a Monitored Retrievable Storage facility, which could provide temporary storage of spent fuel accepted for disposal, and improve the flexibility of the repository development schedule

  12. Criticality safety of high-level tank waste

    International Nuclear Information System (INIS)

    Rogers, C.A.

    1995-01-01

    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

  13. Nuclear waste. DOE's program to prepare high-level radioactive waste for final disposal

    International Nuclear Information System (INIS)

    Bannerman, Carl J.; Owens, Ronald M.; Dowd, Leonard L.; Herndobler, Christopher S.; Purvine, Nancy R.; Stenersen, Stanley G.

    1989-11-01

    In summary, as of December 1988, the four sites collectively stored about 95 million gallons of high-level waste in underground tanks and bins. Approximately 57 million gallons are stored at Hanford, 34 million gallons at Savannah River, 3 million gallons at INEL, and 6 million gallons at West Valley. The waste is in several forms, including liquid, sludge, and dry granular materials, that make it unsuitable for permanent storage in its current state at these locations. Leaks from the tanks, designed for temporary storage, can pose an environmental hazard to surrounding land and water for thousands of years. DOE expects that when its waste processes at Savannah River, West Valley, and Hanford become operational, the high-level radioactive waste stored at these sites will be blended with other materials to immobilize it by forming a glass-like substance. The glass form will minimize the risk of environmental damage and make the waste more acceptable for permanent disposal in a geologic repository. At INEL, DOE is still considering various other immobilization and permanent disposal approaches. In July 1989, DOE estimated that it would cost about $13 billion (in fiscal year 1988 dollars) to retrieve, process, immobilize, and store the high-level waste until it can be moved to a permanent disposal site: about $5.3 billion is expected to be spent at Savannah River, $0.9 billion at West Valley, $2.8 billion at Hanford, and $4.0 billion at INEL. DOE has started construction at Savannah River and West Valley for facilities that will be used to transform the waste into glass (a process known as vitrification). These sites have each encountered schedule delays, and one has encountered a significant cost increase over earlier estimates. More specifically, the Savannah River facility is scheduled to begin high-level waste vitrification in 1992; the West Valley project, based on a January 1989 estimate, is scheduled to begin high-level waste vitrification in 1996, about 8

  14. High-level radioactive waste repositories site selection plan

    International Nuclear Information System (INIS)

    Castanon, A.; Recreo, F.

    1985-01-01

    A general vision of the high level nuclear waste (HLNW) and/or nuclear spent fuel facilities site selection processes is given, according to the main international nuclear safety regulatory organisms quidelines and the experience from those countries which have reached a larger development of their national nuclear programs. (author)

  15. The IAEA's high level radioactive waste management programme

    International Nuclear Information System (INIS)

    Saire, D.E.

    1994-01-01

    This paper presents the different activities that are performed under the International Atomic Energy Agency's (IAEA) high level radioactive waste management programme. The Agency's programme is composed of five main activities (information exchange, international safety standards, R ampersand D activities, advisory services and special projects) which are described in the paper. Special emphasis is placed on the RADioactive WAste Safety Standards (RADWASS) programme which was implemented in 1991 to document international consensus that exists on the safe management of radioactive waste. The paper also raises the question about the need for regional repositories to serve certain countries that do not have the resources or infrastructure to construct a national repository

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

    International Nuclear Information System (INIS)

    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

  17. Development and evaluation of candidate high-level waste forms

    International Nuclear Information System (INIS)

    Bernadzikowski, T.A.

    1981-01-01

    Some seventeen candidate waste forms have been investigated under US Department of Energy programs as potential media for the immobilization and geologic disposal of the high-level radioactive wastes (HLW) resulting from chemical processing of nuclear reactor fuels and targets. Two of these HLW forms were selected at the end of fiscal year (FY) 1981 for intensive development if FY 1982 to 1983. Borosilicate glass was continued as the reference form. A crystalline ceramic waste form, SYNROC, was selected for further product formulation and process development as the alternative to borosilicate glass. This paper describes the bases on which this decision was made

  18. Spanish high level radioactive waste management system issues

    International Nuclear Information System (INIS)

    Espejo, J.M.; Beceiro, A.R.

    1992-01-01

    The Empresa Nacional de Residuos Radiactivos, S.A. (ENRESA) has been limited liability company to be responsible for the management of all kind of radioactive wastes in Spain. This paper provides an overview of the strategy and main lines of action stated in the third General Radioactive Waste Plan, currently in force, for the management of spent nuclear fuel and high - level wastes, as well as an outline of the main related projects, either being developed or foreseen. Aspects concerning the organizational structure, the economic and financing system and the international cooperation are also included

  19. Spanish high level radioactive waste management system issues

    International Nuclear Information System (INIS)

    Ulibarri, A.; Veganzones, A.

    1993-01-01

    The Empresa Nacional de Residuous Radiactivos, S.A. (ENRESA) was set up in 1984 as a state-owned limited liability company to be responsible for the management of all kinds of radioactive wastes in Spain. This paper provides an overview of the strategy and main lines of action stated in the third General Radioactive Waste Plan, currently in force, for the management of spent nuclear fuel and high-level wastes, as well as an outline of the main related projects, either being developed or foreseen. Aspects concerning the organizational structure, the economic and financing system and the international co-operational are also included

  20. High-level nuclear waste disposal: Ethical considerations

    International Nuclear Information System (INIS)

    Maxey, M.N.

    1985-01-01

    Popular skepticism about, and moral objections to, recent legislation providing for the management and permanent disposal of high-level radioactive wastes have derived their credibility from two major sources: government procrastination in enacting waste disposal program, reinforcing public perceptions of their unprecedented danger and the inflated rhetoric and pretensions to professional omnicompetence of influential scientists with nuclear expertise. Ethical considerations not only can but must provide a mediating framework for the resolution of such a polarized political controversy. Implicit in moral objections to proposals for permanent nuclear waste disposal are concerns about three ethical principles: fairness to individuals, equitable protection among diverse social groups, and informed consent through due process and participation

  1. Canadian high-level radioactive waste management system issues

    International Nuclear Information System (INIS)

    Allan, C.J.; Gray, B.R.

    1992-01-01

    In Canada responsibility for the management of radioactive wastes rests with the producer of those wastes. This fundamental principle applies to such diverse wastes as uranium mine and mill tailings, low-level wastes from universities and hospitals, wastes produced at nuclear research establishments, and wastes produced at nuclear generating stations. The federal government has accepted responsibility for historical wastes for which the original producer can no longer be held accountable. Management of radioactive wastes is subject to the regulatory control of the Atomic Energy Control Board, the federal agency responsible for regulating the nuclear industry. In this paper the authors summarize the current situation concerning the management of high level (used nuclear fuel) wastes. In 1981 the two governments also announced that selection of a disposal site would not proceed, and responsibility for site selection and operation would not be assigned until the Concept for used fuel disposal had been reviewed and assessed. Thus the concept assessment is generic rather than site specific. The Concept that has been developed has been designed to conform with safety and performance criteria established by the Atomic Energy Control Board. It is based on burial deep in plutonic rock of the Canadian Shield, using a multi-barrier approach with a series of engineered and natural barriers: these include the waste form, container, buffer and backfill, and the host rock

  2. Predisposal management of high level radioactive waste. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    Radioactive waste is generated in the generation of electricity in nuclear power plants and in the use of radioactive material in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized. The principles and requirements that govern the safety of the management of radioactive waste are presented in 'The Principles of Radioactive Waste Management', 'Legal and Governmental Infrastructure for Nuclear, Radiation, Radioactive Waste and Transport Safety' and 'Predisposal Management of Radioactive Waste, Including Decommissioning'. The objective of this Safety Guide is to provide regulatory bodies and the operators that generate and manage radioactive waste with recommendations on how to meet the principles and requirements established in Refs for the predisposal management of HLW. This Safety Guide applies to the predisposal management of HLW. For liquid HLW arising from the reprocessing of spent fuel the recommendations of this Safety Guide apply from when liquid waste from the first extraction process is collected for storage and subsequent processing. Recommendations and guidance on the storage of spent fuel, whether or not declared as waste, subsequent to its removal from the storage facility of a reactor are provided in Refs. For spent fuel declared as waste this Safety Guide applies to all activities subsequent to its removal from the storage facility of a reactor and prior to its disposal. Requirements pertaining to the transport of spent fuel, whether or not declared as waste, and of all forms of HLW are established. This Safety Guide provides recommendations on the safety aspects of managing HLW, including the planning, design, construction, commissioning, operation and decommissioning of equipment or facilities for the predisposal management of HLW. It addresses the following elements: (a) The characterization and processing (i.e. pretreatment

  3. Glass-solidification method for high level radioactive liquid waste

    International Nuclear Information System (INIS)

    Kawamura, Kazuhiro; Kometani, Masayuki; Sasage, Ken-ichi.

    1996-01-01

    High level liquid wastes are removed with precipitates mainly comprising Mo and Zr, thereafter, the high level liquid wastes are mixed with a glass raw material comprising a composition having a B 2 O 3 /SiO 2 ratio of not less than 0.41, a ZnO/Li 2 O ratio of not less than 1.00, and an Al 2 O 3 /Li 2 O ratio of not less than 2.58, and they are melted and solidified into glass-solidification products. The liquid waste content in the glass-solidification products can be increased up to about 45% by using the glass raw material having such a predetermined composition. In addition, deposition of a yellow phase does not occur, and a leaching rate identical with that in a conventional case can be maintained. (T.M.)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Ericsson, A.M.

    1977-11-01

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

  6. Remote ignitability analysis of high-level radioactive waste

    International Nuclear Information System (INIS)

    Lundholm, C.W.; Morgan, J.M.; Shurtliff, R.M.; Trejo, L.E.

    1992-09-01

    The Idaho Chemical Processing Plant (ICPP), was used to reprocess nuclear fuel from government owned reactors to recover the unused uranium-235. These processes generated highly radioactive liquid wastes which are stored in large underground tanks prior to being calcined into a granular solid. The Resource Conservation and Recovery Act (RCRA) and state/federal clean air statutes require waste characterization of these high level radioactive wastes for regulatory permitting and waste treatment purposes. The determination of the characteristic of ignitability is part of the required analyses prior to calcination and waste treatment. To perform this analysis in a radiologically safe manner, a remoted instrument was needed. The remote ignitability Method and Instrument will meet the 60 deg. C. requirement as prescribed for the ignitability in method 1020 of SW-846. The method for remote use will be equivalent to method 1020 of SW-846

  7. Vitrification of high-level radioactive and hazardous wastes

    International Nuclear Information System (INIS)

    Lutze, W.

    1993-12-01

    The main objective is to summarize work conducted on glasses as waste forms for high-level radioactive fission product solutions up to the late 1980's (section I and II). Section III addresses the question, whether waste forms designed for the immobilization of radioactive residues can be used for the same purpose for hazardous wastes. Of particular interest are those types of hazardous wastes, e.g., fly ashes from municipal combustion plants, easy to convert into glasses or ceramic materials. A large number of base glass compositions has been studied to vitrify waste from reprocessing but only borosilicate glasses with melting temperatures between 1100 C and 1200 C and very good hydrolytic stability is used today. (orig./HP) [de

  8. High-level waste management technology program plan

    International Nuclear Information System (INIS)

    Harmon, H.D.

    1995-01-01

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

  9. High-level waste management technology program plan

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, H.D.

    1995-01-01

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

  10. Determination of performance criteria for high-level solidified nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Heckman, R.A.; Holdsworth, T.

    1979-05-07

    To minimize radiological risk from the operation of a waste management system, performance limits on volatilization, particulate dispersion, and dissolution characteristics of solidified high level waste must be specified. The results show clearly that the pre-emplacement environs are more limiting in establishing the waste form performance criteria than the post-emplacement environs. Absolute values of expected risk are very sensitive to modeling assumptions. The transportation and interim storage operations appear to be most limiting in determining the performance characteristics required. The expected values of risk do not rely upon the repositories remaining intact over the potentially hazardous lifetime of the waste.

  11. Determination of performance criteria for high-level solidified nuclear waste

    International Nuclear Information System (INIS)

    Heckman, R.A.; Holdsworth, T.

    1979-01-01

    To minimize radiological risk from the operation of a waste management system, performance limits on volatilization, particulate dispersion, and dissolution characteristics of solidified high level waste must be specified. The results show clearly that the pre-emplacement environs are more limiting in establishing the waste form performance criteria than the post-emplacement environs. Absolute values of expected risk are very sensitive to modeling assumptions. The transportation and interim storage operations appear to be most limiting in determining the performance characteristics required. The expected values of risk do not rely upon the repositories remaining intact over the potentially hazardous lifetime of the waste

  12. Defense High Level Waste Disposal Container System Description Document

    International Nuclear Information System (INIS)

    2000-01-01

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms (IPWF)) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. 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

  13. Symposium on the development of nuclear waste policy: Siting the high-level nuclear waste repository

    International Nuclear Information System (INIS)

    Pijawka, K.D.; Mushkatel, A.H.

    1991-01-01

    The Nuclear Waste Policy Act of 1982 (NWPA) attempted to formulate a viable national policy for managing the disposal of high-level nuclear wastes. The NWPA authorized the selection of two repository sites: the first to be constructed in the West and a second site developed in the eastern United States. A detailed process for site selection was outlined in the NWPA. In addition, the NWPA authorized open-quotes the development of a waste transportation system; required the Department of Energy (DOE) to submit a proposal to construct a facility for monitored retrievable storage (MRS) after conducting a study of the need for, and feasibility of such a facility; and required the President to evaluate the use of the repositories ... for the disposal of high-level waste resulting from defense activitiesclose quotes (DOE, 1988, p. 1). A series of provisions granting oversight participation to states and Indian tribes, as well as a compensation package for the ultimate host state were also included. Responsibility for implementing the NWPA was assigned to DOE

  14. High Level Waste (HLW) Processing Experience with Increased Waste Loading

    International Nuclear Information System (INIS)

    JANTZEN, CAROL

    2004-01-01

    The Defense Waste Processing Facility (DWPF) Engineering requested characterization of glass samples that were taken after the second melter had been operational for about 5 months. After the new melter had been installed, the waste loading had been increased to about 38 weight percentage after a new quasicrystalline liquidus model had been implemented. The DWPF had also switched from processing with refractory Frit 200 to a more fluid Frit 320. The samples were taken after DWPF observed very rapid buildup of deposits in the upper pour spout bore and on the pour spout insert while processing the high waste loading feedstock. These samples were evaluated using various analytical techniques to determine the cause of the crystallization. The pour stream sample was homogeneous, amorphous, and representative of the feed batch from which it was derived. Chemical analysis of the pour stream sample indicated that a waste loading of 38.5 weight per cent had been achieved. The data analysis indicated that surface crystallization, induced by temperature and oxygen fugacity gradients in the pour spout, caused surface crystallization to occur in the spout and on the insert at the higher waste loadings even though there was no crystallization in the pour stream

  15. Safe immobilization of high-level nuclear reactor wastes

    International Nuclear Information System (INIS)

    Ringwood, A.; Kesson, S.; Ware, N.; Hibberson, W.; Major, A.

    1979-01-01

    The advantages and disadvantages of methods of immobilizing high-level radioactive wastes are discussed. Problems include the devitrification of glasses and the occurrence of radiation damage. An alternative method of radioctive waste immobilization is described in which the waste is incorporated in the constituent minerals of a synthetic rock, Synroc. Synroc is immune from devitrification and is composed of phases which possess crystal structures identical to those of minerals which are known to have retained radioactive elements in geological environments at elevated pressures and tempertures for long periods. The composition and mineralogy of Synroc is given and the process of immobilizing wastes in Synroc is described. Accelerated leaching tests at elevated pressures and temperatures are also described

  16. High-level waste description, inventory and hazard

    International Nuclear Information System (INIS)

    Crandall, J.; Hennelly, E.J.; McElroy, J.L.

    1983-01-01

    High-level nuclear waste (HLW), including its origin, is described and the current differences in definitions discussed. Quantities of defense and commercial radioactive HLW, both volume and curie content, are given. Current waste handling, which is interimin nature, is described for the several sites. The HLW hazard is defined by the times during which various radionuclides are the dominant contributors. The hazard is also compared to that of the ore. Using ICRP-2, which is the legal reference in the US, the hazard of the waste reduces to a level equal to the ore in about 300 years. The disposal plans are summarized and it is shown that regulatory requirements will probably govern disposal operations in such a conservative manner that the risk (product of hazard times probability of release) may well be lower than for any other wastes in existence or perhaps lower than those for any other human endeavor

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

    International Nuclear Information System (INIS)

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

    1996-04-01

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

  18. The development of a high level radioactive waste management strategy

    International Nuclear Information System (INIS)

    Beale, H.

    1979-11-01

    The management of high level radioactive waste, from the removal of spent fuel from reactors to final disposal of vitrified waste, involves a complex choice of operational variables which interact one with another. If the various operations are designed and developed in isolation it will almost certainly lead to suboptimal choice. Management of highly active waste should therefore be viewed as a complete system and analysed in such a way that account is taken of the interactions between the various operations. This system must have clearly defined and agreed objectives as well as criteria against which performance can be judged. A thorough analysis of the system will provide a framework within which the necessary research and development can be carried out in a co-ordinated fashion and lead to an optimised strategy for managing highly active wastes. (author)

  19. Immobilization of high-level wastes into sintered glass: 1

    International Nuclear Information System (INIS)

    Russo, D.O.; Messi de Bernasconi, N.; Audero, M.A.

    1987-01-01

    In order to immobilize the high-level radioactive wastes from fuel elements reprocessing, borosilicate glass was adopted. Sintering experiments are described with the variety VG 98/12 (SiO 2 , TiO 2 , Al 2 O 3 , B 2 O 3 , MgO, CaO and Na 2 O) (which does not present devitrification problems) mixed with simulated calcinated wastes. The hot pressing line (sintering under pressure) was explored in two variants 1: In can; 2: In graphite matrix with sintered pellet extraction. With scanning electron microscopy it is observed that the simulated wastes do not disolve in the vitreous matrix, but they remain dispersed in the same. The results obtained point out that the leaching velocities are independent from the density and from the matrix type employed, as well as from the fact that the wastes do no dissolve in the matrix. (M.E.L.) [es

  20. High-level radioactive waste in Canada. Background paper

    International Nuclear Information System (INIS)

    Fawcett, R.

    1993-11-01

    The disposal of radioactive waste is one of the most challenging environmental problems facing Canada today. Since the Second World War, when Canadian scientists first started to investigate nuclear reactions, there has been a steady accumulation of such waste. Research reactors built in the early postwar years produced small amounts of radioactive material but the volume grew steadily as the nuclear power reactors constructed during the 1960s and 1970s began to spawn used fuel bundles. Although this radioactive refuse has been safely stored for the short term, no permanent disposal system has yet been fully developed and implemented. Canada is not alone in this regard. A large number of countries use nuclear power reactors but none has yet put in place a method for the long-term disposal of the radioactive waste. Scientists and engineers throughout the world are investigating different possibilities; however, enormous difficulties remain. In Canada, used fuel bundles from nuclear reactors are defined as high-level waste; all other waste created at different stages in the nuclear fuel cycle is classified as low-level. Although disposal of low-level waste is an important issue, it is a more tractable problem than the disposal of high-level waste, on which this paper will concentrate. The paper discusses the nuclear fuel waste management program in Canada, where a long-term disposal plan has been under development by scientists and engineers over the past 15 years, but will not be completed for some time. Also discussed are responses to the program by parliamentary committees and aboriginal and environmental groups, and the work in the area being conducted in other countries. (author). 1 tab

  1. High-level radioactive waste in Canada. Background paper

    Energy Technology Data Exchange (ETDEWEB)

    Fawcett, R [Library of Parliament, Ottawa, ON (Canada). Science and Technology Div.

    1993-11-01

    The disposal of radioactive waste is one of the most challenging environmental problems facing Canada today. Since the Second World War, when Canadian scientists first started to investigate nuclear reactions, there has been a steady accumulation of such waste. Research reactors built in the early postwar years produced small amounts of radioactive material but the volume grew steadily as the nuclear power reactors constructed during the 1960s and 1970s began to spawn used fuel bundles. Although this radioactive refuse has been safely stored for the short term, no permanent disposal system has yet been fully developed and implemented. Canada is not alone in this regard. A large number of countries use nuclear power reactors but none has yet put in place a method for the long-term disposal of the radioactive waste. Scientists and engineers throughout the world are investigating different possibilities; however, enormous difficulties remain. In Canada, used fuel bundles from nuclear reactors are defined as high-level waste; all other waste created at different stages in the nuclear fuel cycle is classified as low-level. Although disposal of low-level waste is an important issue, it is a more tractable problem than the disposal of high-level waste, on which this paper will concentrate. The paper discusses the nuclear fuel waste management program in Canada, where a long-term disposal plan has been under development by scientists and engineers over the past 15 years, but will not be completed for some time. Also discussed are responses to the program by parliamentary committees and aboriginal and environmental groups, and the work in the area being conducted in other countries. (author). 1 tab.

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

    International Nuclear Information System (INIS)

    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

  3. Immobilisation of high level nuclear reactor wastes in SYNROC

    Energy Technology Data Exchange (ETDEWEB)

    Ringwood, A E; Kesson, S E; Ware, N G; Hibberson, W; Major, A [Australian National Univ., Canberra. Inst. of Advanced Studies

    1979-03-15

    It is stated that the elements occurring in high-level nuclear reactor wastes can be safely immobilised by incorporating them within the crystal lattices of the constituent minerals of a synthetic rock (SYNROC). The preferred form of SYNROC can accept up to 20% of high level waste calcine to form dilute solid solutions. The constituent minerals, or close structural analogues, have survived in a wide range of geochemical environments for periods of 20 to 2,000 Myr whilst immobilising the same elements present in nuclear wastes. SYNROC is unaffected by leaching for 24 hours in pure water or 10 wt % NaCl solution at high temperatures and pressure whereas borosilicate glasses completely decompose in a few hours in much less severe hydrothermal conditions. The combination of these leaching results with the geological evidence of long-term stability indicates that SYNROC would be vastly superior to glass in its capacity to safely immobilise nuclear wastes, when buried in a suitable geological repository. A dense, compact, mechanically strong form of SYNROC suitable for geological disposal can be produced by a process as economical as that which incorporates radioactive waste in borosilicate glasses.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  7. Safe disposal of high-level radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Ringwood, A E [Australian National Univ., Canberra. Research School of Earth Sciences

    1980-10-01

    Current strategies in most countries favour the immobilisation of high-level radioactive wastes in borosilicate glasses, and their burial in large, centralised, mined repositories. Strong public opposition has been encountered because of concerns over safety and socio-political issues. The author develops a new disposal strategy, based on immobilisation of wastes in an extremely resistant ceramic, SYNROC, combined with burial in an array of widely dispersed, very deep drill holes. It is demonstrated that the difficulties encountered by conventional disposal strategies can be overcome by this new approach.

  8. Study on high-level waste geological disposal metadata model

    International Nuclear Information System (INIS)

    Ding Xiaobin; Wang Changhong; Zhu Hehua; Li Xiaojun

    2008-01-01

    This paper expatiated the concept of metadata and its researches within china and abroad, then explain why start the study on the metadata model of high-level nuclear waste deep geological disposal project. As reference to GML, the author first set up DML under the framework of digital underground space engineering. Based on DML, a standardized metadata employed in high-level nuclear waste deep geological disposal project is presented. Then, a Metadata Model with the utilization of internet is put forward. With the standardized data and CSW services, this model may solve the problem in the data sharing and exchanging of different data form A metadata editor is build up in order to search and maintain metadata based on this model. (authors)

  9. Safety of geologic disposal of high level radioactive waste

    International Nuclear Information System (INIS)

    Zaitsu, Tomohisa; Ishiguro, Katsuhiko; Masuda, Sumio

    1992-01-01

    This article introduces current concepts of geologic disposal of high level radioactive waste and its safety. High level radioactive waste is physically stabilized by solidifying it in a glass form. Characteristics of deep geologic layer are presented from the viewpoint of geologic disposal. Reconstruction of multi-barrier system receives much attention to secure the safety of geologic disposal. It is important to research performance assessment of multi-barrier system for preventing dissolution or transfer of radionuclides into the ground water. Physical and chemical modeling for the performance assessment is outlined in the following terms: (1) chemical property of deep ground water, (2) geochemical modeling of artificial barrier spatial water, (3) hydrology of deep ground water, (4) hydrology of the inside of artificial barrier, and (5) modeling of radionuclide transfer from artificial barrier. (N.K.)

  10. Leaching behavior of simulated high-level waste glass

    International Nuclear Information System (INIS)

    Kamizono, Hiroshi

    1987-03-01

    The author's work in the study on the leaching behavior of simulated high-level waste (HLW) glass were summarized. The subjects described are (1) leach rates at high temperatures, (2) effects of cracks on leach rates, (3) effects of flow rate on leach rates, and (4) an in-situ burial test in natural groundwater. In the following section, the leach rates obtained by various experiments were summarized and discussed. (author)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Peterson, P.F.

    1999-01-01

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

  13. Soil-structure interaction effects on high level waste tanks

    International Nuclear Information System (INIS)

    Miller, C.A.; Costantino, C.J.; Heymsfeld, E.

    1991-01-01

    High Level Waste Tanks consist of steel tanks located in concrete vaults which are usually completely embedded in the soil. Many of these tanks are old and were designed to seismic standards which are not compatible with current requirements. The objective if this paper is to develop simple methods of modeling SSI effects for such structures and to obtain solutions for a range of parameters that can be used to identify significant aspects of the problem

  14. Development of cermets for high-level radioactive waste fixation

    International Nuclear Information System (INIS)

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

    1979-01-01

    A method is currently under development for the solidification and fixation of commercial and defense high-level radioactive wastes in the form of ceramic particles encapsulated by metal, i.e., a cermet. The chemical and physical processing techniques which have been developed and the properties of the resulting cermet bodies are described in this paper. These cermets have the advantages of high thermal conductivity and low leach rates

  15. Research on high level radioactive waste repository seismic design criteria

    International Nuclear Information System (INIS)

    Jing Xu

    2012-01-01

    Review seismic hazard analysis principle and method in site suitable assessment process of Yucca Mountain Project, and seismic design criteria and seismic design basis in primary design process. Demonstrated spatial character of seismic hazard by calculated regional seismic hazard map. Contrasted different level seismic design basis to show their differences and relation. Discussed seismic design criteria for preclosure phrase of high level waste repository and preference goal under beyond design basis ground motion. (author)

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

    International Nuclear Information System (INIS)

    Johnson, P.E.; Joy, D.S.; Pope, R.B.; Thomas, T.M.; Lester, P.B.

    1994-01-01

    The TRANSCOM (transportation tracking and communication) system is the US 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 FY 1993 to track almost 100 shipments within the US DOE complex, and it is accessed weekly by 10 to 20 users

  17. Apparatus for Crossflow Filtration Testing of High Level Waste Samples

    International Nuclear Information System (INIS)

    Nash, C.

    1998-05-01

    Remotely-operated experimental apparatuses for verifying crossflow filtration of high level nuclear waste have been constructed at the Savannah River Site (SRS). These units have been used to demonstrate filtration processes at the Savannah River Site, Oak Ridge National Laboratory, the Idaho National Engineering and Environmental Laboratory, and Pacific Northwest National Laboratory. The current work covers the design considerations for experimentation as well as providing results from testing at SRS

  18. Strategic lessons in high-level waste management planning

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, Neil

    1999-07-01

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

  19. Strategic lessons in high-level waste management planning

    International Nuclear Information System (INIS)

    Chapman, Neil

    1999-01-01

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

  20. Potential host media for a high-level waste repository

    Energy Technology Data Exchange (ETDEWEB)

    Hustrulid, W

    1982-01-01

    Earlier studies of burial of radioactive wastes in geologic repositories had concentrated on salt formations for well-publicized reasons. However, under the Carter administration, significant changes were made in the US nuclear waste management program. Changes which were made were: (1) expansion of the number of rock types under consideration; (2) adoption of the multiple-barrier approach to waste containment; (3) additional requirements for waste retrieval; and (4) new criteria proposed by the Nuclear Regulatory Commission for the isolation of high-level waste in geologic repositories. Results of the studies of different types of rocks as repository sites are summarized herein. It is concluded that each generic rock type has certain advantages and disadvantages when considered from various aspects of the waste disposal problem and that characteristics of rocks are so varied that a most favorable or least favorable rock type cannot be easily identified. This lack of definitive characteristics of rocks makes site selection and good engineering barriers very important for containment of the wastes. (BLM)

  1. High level radioactive wastes: Considerations on final disposal

    International Nuclear Information System (INIS)

    Ciallella, Norberto R.

    2000-01-01

    When at the beginnings of the decade of the 80 the National Commission on Atomic Energy (CNEA) in Argentina decided to study the destination of the high level radioactive wastes, was began many investigations, analysis and multidisciplinary evaluations that be origin to a study of characteristics never before carried out in Argentina. For the first time in the country was faced the study of an environmental eventual problem, several decades before that the problem was presented. The elimination of the high level radioactive wastes in the technological aspects was taken in advance, avoiding to transfer the problems to the future generations. The decision was based, not only in technical evaluations but also in ethical premises, since it was considered that the future generations may enjoy the benefits of the nuclear energy and not should be solve the problem. The CNEA in Argentina in 1980 decided to begin a feasibility study and preliminary engineering project for the construction of the final disposal of high level radioactive wastes

  2. Progress report on safety research of high-level waste management for the period April, 1981 to March, 1982

    International Nuclear Information System (INIS)

    Tashiro, Shingo

    1982-10-01

    Main results obtained on Safety Research of High-Level Waste Management in 1981 were edited. The research tjeme are following. (1) Characterization of vitrified waste. (2) Alternative waste form development. (3) Durability tests for HLW storage facility. (4) Safety evaluation of geologic disposal. (5) Preparation for hot test. (author)

  3. Long-lived legacy: Managing high-level and transuranic waste at the DOE Nuclear Weapons Complex. Background paper

    International Nuclear Information System (INIS)

    1991-05-01

    The document focuses on high-level and transuranic waste at the DOE nuclear weapons complex. Reviews some of the critical areas and aspects of the DOE waste problem in order to provide data and further analysis of important issues. Partial contents, High-Level Waste Management at the DOE Weapons Complex, are as follows: High-Level Waste Management: Present and Planned; Amount and Distribution; Current and Potential Problems; Vitrification; Calcination; Alternative Waste Forms for the Idaho National Engineering Laboratory; Technologies for Pretreatment of High-Level Waste; Waste Minimization; Regulatory Framework; Definition of High-Level Waste; Repository Delays and Contingency Planning; Urgency of High-Level Tank Waste Treatment; Technologies for High-Level Waste Treatment; Rethinking the Waste Form and Package; Waste Form for the Idaho National Engineering Laboratory; Releases to the Atmosphere; Future of the PUREX Plant at Hanford; Waste Minimization; Tritium Production; International Cooperation; Scenarios for Future HLW Production. Partial contents of Chapter 2, Managing Transuranic Waste at the DOE Nuclear Weapons Complex, are as follows: Transuranic Waste at Department of Energy Sites; Amount and Distribution; Waste Management: Present and Planned; Current and Potential Problems; Three Technologies for Treating Retrievably Stored Transuranic Waste; In Situ Vitrification; The Applied Research, Development, Demonstration, Testing, and Evaluation Plan (RDDT ampersand E); Actinide Conversion (Transmutation); Waste Minimization; The Regulatory Framework; Definition of, and Standards for, Disposal of Transuranic Waste; Repository Delays; Alternative Storage and Disposal Strategies; Remediation of Buried Waste; The Waste Isolation Pilot Plant; Waste Minimization; Scenarios for Future Transuranic Waste Production; Conditions of No-Migration Determination

  4. Post treatment of high-level nuclear fuel wastes

    International Nuclear Information System (INIS)

    Berreth, J.R.; Cole, H.S.; Hoskins, A.P.; Lewis, L.C.; Samsel, E.G.

    1975-01-01

    The glass-ceramic product prepared from fluidized-bed calcined synthetic commercial wastes, based on data obtained to date, has many of the properties desired for long-term storage. Although more characterization is necessitated, the product's high-calcine content will decrease the number of storage canisters required and use a minimum of product-forming additives, resulting in significant process cost savings. The product remains in a solid, nonflowing form at temperatures close to the preparation temperature and yet is prepared at relatively low temperatures. The product has void spaces to accommodate radiolytic gas formation, but is hard and dense and has very low leach rates. Process features, such as no direct product contact with furnace or storage canisters, will minimize corrosion of both process equipment and storage canisters

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

    International Nuclear Information System (INIS)

    Decker, L.A.

    1998-04-01

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

  6. High level waste forms: glass marbles and thermal spray coatings

    International Nuclear Information System (INIS)

    Treat, R.L.; Oma, K.H.; Slate, S.C.

    1982-01-01

    A process that converts high-level waste to glass marbles and then coats the marbles has been developed at Pacific Northwest Laboratory (PNL) under sponsorship of the US Department of Energy. The process consists of a joule-heated glass melter, a marble-making device based on a patent issued to Corning Glass Works, and a coating system that includes a plasma spray coater and a marble tumbler. The process was developed under the Alternative Waste Forms Program which strived to improve upon monolithic glass for immobilizing high-level wastes. Coated glass marbles were found to be more leach-resistant, and the marbles, before coating were found to be very homogeneous, highly impact resistant, and conductive to encapsulation in a metal matric for improved heat transfer and containment. Marbles are also ideally suited for quality assurance and recycling. However, the marble process is more complex, and marbles require a larger number of canisters for waste containment and have a higher surface area than do glass monoliths

  7. Pelleted waste form for high-level ICPP wastes

    International Nuclear Information System (INIS)

    Lamb, K.M.; Priebe, S.J.; Cole, H.S.; Taki, B.D.

    1979-01-01

    Simulated zirconia type calcined waste is pelletized on a 41-cm dia disc pelletizer using 5% bentonite, 2% metakaolin, and 2% boric acid as a solid binder and 7M phosphoric plus 4M nitric acid as a liquid binder. After heat treatment at 800 0 C for 2 hours, the pellets are impact resistant and have a leach resistance of 10 -4 g/cm 2 /day, based on Soxhlet leaching for 100 hours at 95 0 C with distilled water. An integrated pilot plant is being fabricated to verify the process. 1 figure, 4 tables

  8. Pelleted waste form for high-level ICPP wastes

    International Nuclear Information System (INIS)

    Lamb, K.M.; Priebe, S.J.; Cole, H.S.; Taki, B.d.

    1979-01-01

    Simulated zirconia-type calcined waste is pelletized on a 41-cm diameter disc pelletizer using 5% bentonite, 2% metakaolin, and 2% boric acid as a solid binder and 7M phosphoric plus 4M nitric acid as a liquid binder. After heat treatment at 800 0 C for 2 hours the pellets are impact resistant and have a leach resistance of 10 -4 g/cm 2 . day, based on Soxhlet leaching for 100 hours at 95 0 C with distilled water. An integrated pilot plant is being fabricated to verify the process. 1 figure, 4 tables

  9. Separation processes for high-level radioactive waste treatment

    International Nuclear Information System (INIS)

    Sutherland, D.G.

    1992-11-01

    During World War II, production of nuclear materials in the United States for national defense, high-level waste (HLW) was generated as a byproduct. Since that time, further quantities of HLW radionuclides have been generated by continued nuclear materials production, research, and the commercial nuclear power program. In this paper HLW is defined as the highly radioactive material resulting from the processing of spent nuclear fuel. The HLW is the liquid waste generated during the recovery of uranium and plutonium in a fuel processing plant that generally contains more than 99% of the nonvolatile fission products produced during reactor operation. Since this paper deals with waste separation processes, spent reactor fuel elements that have not been dissolved and further processed are excluded

  10. Fluidized bed system for calcination of high level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Pande, D P; Prasad, T L; Yadgiri, N K; Theyyunni, T K [Process Engineering and Systems Development Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    During the operation of nuclear facilities significant quantities of radiochemical liquid effluents of different concentrations and varying chemical compositions are generated. These effluents contain activated radionuclides, corrosion products and fission products. The advantage of feeding the waste in solid form into the vitrifying equipment are multifold. Efforts are therefore made in many countries to calcine the high level waste, and obtain waste in the oxide form before the same is mixed with glass forming additives and fed into the melter unit. An experimental rig for fluidized bed calcination is constructed for carrying out the detailed investigation of this process, in order to adopt the same for plant scale application. To achieve better gas-solid contact and avoid raining down of solids, a distributor of bubble cap type was designed. A review of existing experience at various laboratories and design of new experimental facility for development of calciners are given. (author). 11 refs., 5 figs.

  11. A comparison of high-level waste form characteristics

    International Nuclear Information System (INIS)

    Salmon, R.; Notz, K.J.

    1991-01-01

    The US DOE is responsible for the eventual disposal in a repository of spent fuels, high-level waste (HLW) and other radioactive wastes that may require long-term isolation. This includes light-water reactor (LWR) spent fuel and immobilized HLW as the two major sources, plus other forms including non-LWR spent fuels and miscellaneous sources (such as activated metals in the Greater-Than-Class-C category). The Characteristics Data Base, sponsored by DOE's Office of Civilian Radioactive Waste Management (OCRWM), was created to systematically tabulate the technical characteristics of these materials. Data are presented here on the immobilized HLW forms that are expected to be produced between now and 2020

  12. Coupled processes in NRC high-level waste research

    International Nuclear Information System (INIS)

    Costanzi, F.A.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Valentine, J.H.; 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

  14. EPA's high-level waste standards and waste package performance

    International Nuclear Information System (INIS)

    Meyers, S.

    1985-01-01

    The seven assurance requirements EPA was considering were these: (1) Disposal systems shall not depend on active institutional controls for more than 100 years after disposal; (2) Long-term disposal system performance should be monitored for a reasonable time as a supplement to other types of protection; (3) Disposal systems shall be marked and their locations recorded in all appropriate government records; (4) Disposal systems shall be designed with several different types of barriers, both natural and engineered; (5) Sites should not be located where scarce or easily accessible resources are located; (6) Site selection should consider the relative isolation offered by potential alternatives; and (7) Wastes shall be recoverable for a reasonable time after disposal. (orig./PW)

  15. Study of heat diffusion in a granitic geologic formation of high level radioactive wastes

    International Nuclear Information System (INIS)

    Goldstein, S.; Juignet, N.

    1980-06-01

    Thermal study of granitic underground storage of vitrified high level radioactive wastes in a regular network of shafts and galleries. The aim is to show influence on temperature rise of the geologic formation of main parameters to define the storage zone and to determine the network dimension in function of the rock properties. Two models were studied allowing a rapid variation of geometrical and physical parameters. A numerical method using finite element method or Green functions were used for calculations. Temperatures are determined either for the whole storage site or a unit cell of the lattice [fr

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    1992-12-01

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

  18. Alternative processes for managing existing commercial high-level radioactive wastes

    International Nuclear Information System (INIS)

    1976-04-01

    A number of alternatives are discussed for managing high-level radioactive waste presently stored at the West Valley, New York, plant owned by Nuclear Fuel Services, Inc. These alternatives (liquid storage, conversion to cement, shale fracturing, shale cement, calcination, aqueous silicate, conversion to glass, and salt cake) are limited to concepts presently under active investigation by ERDA. Each waste management option is described and examined regarding the status of the technology; its applications to managing NFS waste; its advantages and disadvantages; the research and development needed to implement the option; safety considerations; and estimated costs and time to implement the process

  19. High-level waste processing and conditioning: vitrification

    International Nuclear Information System (INIS)

    Bonniaud, R.

    1981-02-01

    The vitrification process used to treat fission product solutions at the Marcoule Vitrification Plant is described. The type of waste processed is characterized by its very high activity and the long lifetimes of some of the emitters that it contains. The performance obtained with this process is given together with the future developments envisaged. The storage of glasses is described as well as their behavior with time [fr

  20. Engineering materials for high level radioactive waste repository

    International Nuclear Information System (INIS)

    Wen Zhijian

    2009-01-01

    Radioactive wastes can arise from a wide range of human activities and have different physical and chemical forms with various radioactivity. The high level radioactive wastes (HLW)are characterized by nuclides of very high initial radioactivity, large thermal emissivity and the long life-term. The HLW disposal is highly concerned by the scientists and the public in the world. At present, the deep geological disposal is regarded as the most reasonable and effective way to safely dispose high-level radioactive wastes in the world. The conceptual model of HLW geological disposal in China is based on a multi-barrier system that combines an isolating geological environment with an engineering barrier system(EBS). The engineering materials in EBS include the vitrified HLW, canister, overpack, buffer materials and backfill materials. Referring to progress in the world, this paper presents the function, the requirement for material selection and design, and main scientific projects of R and D of engineering materials in HLW repository. (authors)

  1. Space augmentation of military high-level waste disposal

    International Nuclear Information System (INIS)

    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 predicability 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

  2. Risk assessments for the disposal of high level radioactive wastes

    International Nuclear Information System (INIS)

    Smith, C.F.

    1975-01-01

    The risks associated with the disposal of high level wastes derive from the potential for release of radioactive materials into the environment. The assessment of these risks requires a methodology for risk analysis, an identification of the radioactive sources, and a method by which to express the relative hazard of the various radionuclides that comprise the high level waste. The development of a methodology for risk analysis is carried out after a review of previous work in the area of probabilistic risk assessment. The methodology suggested involves the probabilistic analysis of a general accident consequence distribution. In this analysis, the frequency aspect of the distribution is treated separately from the normalized probability function. At the final stage of the analysis, the frequency and probability characteristics of the distribution are recombined to provide an estimate of the risk. The characterization of the radioactive source term is accomplished using the ORIGEN computer code. Calculations are carried out for various reactor types and fuel cycles, and the overall waste hazard for a projected thirty-five year nuclear power program is determined

  3. Control of high level radioactive waste-glass melters

    International Nuclear Information System (INIS)

    Bickford, D.F.; Choi, A.S.

    1991-01-01

    Slurry Fed Melters (SFM) are being developed in the United States, Europe and Japan for the conversion of high-level radioactive waste to borosilicate glass for permanent disposal. The high transition metal, noble metal, nitrate, organic, and sulfate contents of these wastes lead to unique melter redox control requirements. Pilot waste-glass melter operations have indicated the possibility of nickel sulfide or noble-metal fission-product accumulation on melter floors, which can lead to distortion of electric heating patterns, and decrease melter life. Sulfide formation is prevented by control of the redox chemistry of the melter feed. The redox state of waste-glass melters is determined by balance between the reducing potential of organic compounds in the feed, and the oxidizing potential of gases above the melt, and nitrates and polyvalent elements in the waste. Semiquantitative models predicting limitations of organic content have been developed based on crucible testing. Computerized thermodynamic computations are being developed to predict the sequence and products of redox reactions and is assessing process variations. Continuous melter test results have been compared to improved computer staged-thermodynamic-models of redox behavior. Feed chemistry control to prevent sulfide and moderate noble metal accumulations are discussed. 17 refs., 3 figs

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

    International Nuclear Information System (INIS)

    Zapp, P.E.

    1994-01-01

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

  5. Strategies for high-level radioactive waste management: the U.S. experience

    International Nuclear Information System (INIS)

    Cotton, T.A.

    1984-01-01

    Technology exists or is under development for the safe, retrievable storage of spent fuel from commercial nuclear reactors and high-level waste from reprocessing that fuel, for many decades, and no insuperable scientific obstacles to permanent disposal of spent fuel or high-level waste in geologic repositories have been identified. However, there are significant institutional obstacles to developing such repositories: strong local opposition to siting and the requirement for a sustained commitment of money and skilled manpower over a period of decades. These create strong incentives to defer the political and economic costs of developing disposal facilities by using less expensive interim storage; yet continued deferral may affect the acceptability of nuclear power. Thus the principal strategic policy issue in high-level waste management is how rapidly to develop disposal facilities. Some countries plan decades of storage before choosing a repository site or a disposal technology, while the United States has enacted a law requiring operation of a geologic repository by 1998. This paper discusses waste management strategic issues and major provisions of the U.S. law, emphasizing those measures dealing with the institutional obstacles to developing geologic repositories. (author)

  6. Proposal for basic safety requirements regarding the disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    1980-04-01

    A working group commissioned to prepare proposals for basic safety requirements for the storage and transport of radioactive waste prepared its report to the Danish Agency of Environmental Protection. The proposals include: radiation protection requirements, requirements concerning the properties of high-level waste units, the geological conditions of the waste disposal location, the supervision of waste disposal areas. The proposed primary requirements for safety evaluation of the disposal of high-level waste in deep geological formations are of a general nature, not being tied to specific assumptions regarding the waste itself, the geological and other conditions at the place of disposal, and the technical methods of disposal. It was impossible to test the proposals for requirements on a working repository. As no country has, to the knowledge of the working group, actually disposed of hifg-level radioactive waste or approved of plans for such disposal. Methods for evaluating the suitability of geological formations for waste disposal, and background material concerning the preparation of these proposals for basic safety requirements relating to radiation, waste handling and geological conditions are reviewed. Appended to the report is a description of the phases of the fuel cycle that are related to the storage of spent fuel and the disposal of high-level reprocessing waste in a salt formation. It should be noted that the proposals of the working group are not limited to the disposal of reprocessed fuel, but also include the direct disposal of spent fuel as well as disposal in geological formations other than salt. (EG)

  7. Effect of canister size on costs of disposal of SRP high-level wastes

    International Nuclear Information System (INIS)

    McDonell, W.R.

    1982-01-01

    The current plan for managing the high-level nuclear wastes at the Savannah River Plant (SRP) calls for processing them into solid forms contained in stainless steel canisters for eventual disposal in a federal geologic repository. A new SRP facility called the Defense Waste Processing Facility (DWPF) is being designed for the onsite waste processing operations. Preliminary evaluations indicate that costs of the overall disposal operation will depend significantly on the size of the canisters, which determines the number of waste forms to be processed. The objective of this study was to evaluate the effects of canister size on costs of DWPF process operations, including canister procurement, waste solidification, and interim storage, on offsite transport, and on repository costs of disposal, including provision of suitable waste packages

  8. High-level waste repository-induced effects

    Energy Technology Data Exchange (ETDEWEB)

    Leupin, O.X.; Marschall, P.; Johnson, L.; Cloet, V.; Schneider, J. [National Cooperative for the Disposal of Radioactive Waste (NAGRA), Wettingen (Switzerland); Smith, P. [Safety Assessment Management Ltd, Henley-On-Thames, Oxfordshire (United Kingdom); Savage, D. [Savage Earth Associates Ltd, Bournemouth, Dorset (United Kingdom); Senger, R. [Intera Inc., Ennetbaden (Switzerland)

    2016-10-15

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

  9. High-level waste repository-induced effects

    International Nuclear Information System (INIS)

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

    2016-10-01

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

  10. Electrochemical probing of high-level radioactive waste tanks containing washed sludge and precipitates

    International Nuclear Information System (INIS)

    Bickford, D.F.; Congdon, J.W.; Oblath, S.B.

    1987-01-01

    At the U.S. Department of Energy's Savannah River Plant, corrosion of carbon steel storage tanks containing alkaline, high-level radioactive waste is controlled by specification of limits on waste composition and temperature. Processes for the preparation of waste for final disposal will result in waste with low corrosion inhibitor concentrations and, in some cases, high aromatic organic concentrations, neither of which are characteristic of previous operations. Laboratory tests, conducted to determine minimum corrosion inhibitor levels indicated pitting of carbon steel near the waterline for proposed storage conditions. In situ electrochemical measurements of full-scale radioactive process demonstrations have been conducted to assess the validity of laboratory tests. Probes included pH, Eh (potential relative to a standard hydrogen electrode), tank potential, and alloy coupons. In situ results are compared to those of the laboratory tests, with particular regard given to simulated solution composition

  11. A critically educated public explores high level radioactive waste management

    International Nuclear Information System (INIS)

    Blum, J.E.

    1994-01-01

    It is vital to the citizens of Nevada that they and their children are given an opportunity to explore all sides of the characterization of Yucca Mountain as a potential repository site for spent nuclear fuel. The state-wide, national and international implications demand a reasoned and complete approach to this issue, which has become emotionally and irrationally charged and fueled by incomplete perception and information. The purpose of this paper is to provide curriculum suggestions and recommend concomitant policy developments that will lead to the implementation of a Critical Thinking (CT) approach to High Level Radioactive Waste Management

  12. Treatment of High-Level Waste Arising from Pyrochemical Processes

    International Nuclear Information System (INIS)

    Lizin, A.A.; Kormilitsyn, M.V.; Osipenko, A.G.; Tomilin, S.V.; Lavrinovich, Yu.G.

    2013-01-01

    JSC “SSC RIAR” has been performing research and development activities in support of closed fuel cycle of fast reactor since the middle of 1960s. Fuel cycle involves fabrication and reprocessing of spent nuclear fuel (SNF) using pyrochemical methods of reprocessing in molten alkali metal chlorides. At present pyrochemical methods of SNF reprocessing in molten chlorides has reached such a level in their development that makes it possible to compare their competitiveness with classic aqueous methods. Their comparative advantage lies in high safety, compactness, high protectability as to nonproliferation of nuclear materials, and reduction of high level waste volume

  13. High-level wastes: DOE names three sites for characterization

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    DOE announced in May 1986 that there will be there site characterization studies made to determine suitability for a high-level radioactive waste repository. The studies will include several test drillings to the proposed disposal depths. Yucca Mountain, Nevada; Deaf Smith Country, Texas, and Hanford, Washington were identified as the study sites, and further studies for a second repository site in the East were postponed. The affected states all filed suits in federal circuit courts because they were given no advance warning of the announcement of their selection or the decision to suspend work on a second repository. Criticisms of the selection process include the narrowing or DOE options

  14. Midwestern High-Level Radioactive Waste Transportation Project

    International Nuclear Information System (INIS)

    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

  15. High level radioactive waste vitrification process equipment component testing

    International Nuclear Information System (INIS)

    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

  16. Powder technological vitrification of simulated high-level waste

    International Nuclear Information System (INIS)

    Gahlert, S.

    1988-03-01

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

  17. Electropolishing decontamination system for high-level waste canisters

    International Nuclear Information System (INIS)

    Larson, D.E.; Berger, D.N.; Allen, R.P.; Bryan, G.H.; Place, B.G.

    1988-10-01

    As part of a US Department of Energy (DOE) project agreement with the Federal Ministry for Research and Technology (BMFT) in the Federal Republic of Germany (FRG). The Nuclear Waste Treatment Program at the Pacific Northwest Laboratory (PNL) is preparing 30 radioactive canisters containing borosilicate glass for use in high-level waste repository related tests at the Asse Salt Mine. After filling, the canisters will be welded closed and decontaminated in preparation for shipping to the FRG. Electropolishing was selected as the primary decontamination approach, and an electropolishing system with associated canister inspection equipment has been designed and fabricated for installation in a large hot cell. This remote electropolishing system, which is currently undergoing preliminary testing, is described in this report. 3 refs., 3 figs., 1 tab

  18. Monitoring of geological repositories for high level radioactive waste

    International Nuclear Information System (INIS)

    2001-04-01

    Geological repositories for disposal of high level radioactive waste are designed to provide isolation of the waste from human environment for many thousands of years. This report discusses the possible purposes for monitoring geological repositories at the different stages of a repository programme, the use that may be made of the information obtained and the techniques that might be applied. This report focuses on the different objectives that monitoring might have at various stages of a programme, from the initiation of work on a candidate site, to the period after repository closure. Each objective may require somewhat different types of information, or may use the same information in different ways. Having evaluated monitoring requirements, the report concludes with a brief evaluation of available monitoring techniques

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

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

    International Nuclear Information System (INIS)

    Mc Menamin, T.

    1989-01-01

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

  2. Storage of radioactive wastes

    International Nuclear Information System (INIS)

    1992-07-01

    Even if the best waste minimization measures are undertaken throughout radioisotope production or usage, significant radioactive wastes arise to make management measures essential. For developing countries with low isotope usage and little or no generation of nuclear materials, it may be possible to handle the generated waste by simply practicing decay storage for several half-lives of the radionuclides involved, followed by discharge or disposal without further processing. For those countries with much larger facilities, longer lived isotopes are produced and used. In this situation, storage is used not only for decay storage but also for in-process retention steps and for the key stage of interim storage of conditioned wastes pending final disposal. The report will serve as a technical manual providing reference material and direct step-by-step know-how to staff in radioisotope user establishments and research centres in the developing Member States without nuclear power generation. Considerations are limited to the simpler storage facilities. The restricted quantities and low activity associated with the relevant wastes will generally permit contact-handling and avoid the need for shielding requirements in the storage facilities or equipment used for handling. A small quantity of wastes from some radioisotope production cells and from reactor cooling water treatment may contain sufficient short lived activity from activated corrosion products to require some separate decay storage before contact-handling is suitable. 16 refs, 12 figs, 8 tabs

  3. Radioactive waste storage issues

    International Nuclear Information System (INIS)

    Kunz, D.E.

    1994-01-01

    In the United States we generate greater than 500 million tons of toxic waste per year which pose a threat to human health and the environment. Some of the most toxic of these wastes are those that are radioactively contaminated. This thesis explores the need for permanent disposal facilities to isolate radioactive waste materials that are being stored temporarily, and therefore potentially unsafely, at generating facilities. Because of current controversies involving the interstate transfer of toxic waste, more states are restricting the flow of wastes into - their borders with the resultant outcome of requiring the management (storage and disposal) of wastes generated solely within a state's boundary to remain there. The purpose of this project is to study nuclear waste storage issues and public perceptions of this important matter. Temporary storage at generating facilities is a cause for safety concerns and underscores, the need for the opening of permanent disposal sites. Political controversies and public concern are forcing states to look within their own borders to find solutions to this difficult problem. Permanent disposal or retrievable storage for radioactive waste may become a necessity in the near future in Colorado. Suitable areas that could support - a nuclear storage/disposal site need to be explored to make certain the health, safety and environment of our citizens now, and that of future generations, will be protected

  4. Managing the nation's high-level radioactive waste: key issues and recommendations

    International Nuclear Information System (INIS)

    1981-07-01

    To date, no unified national plan has been adopted to develop and implement a comprehensive system of management and disposal of high-level radioactive waste in the United States. Growing public concern about this problem has resulted in a number of recent efforts to develop a national high-level waste management policy. The 96th Congress strove to resolve the central issues, but ultimately failed to pass legislation, partly because of disagreements about the appropriate role of states in the siting of repositories for military waste. Outside government, a number of organizations convened representatives of diverse groups concerned with national high-level radioactive waste management to seek agreement on the major elements of national policy. One such organization was RESOLVE, Center for Environmental Conflict Resolution, which in May 1981 was merged into The Conservation Foundation. RESOLVE convened Forum II, a series of discussions among representatives of environmental, industrial, governmental, and citizen interest groups, in 1981 specifically to address the issues blocking Congressional agreement on high-level waste policy. This report contains the recommendations which resulted from these deliberations. Reprocessing, interim storage, respository development, and licensing requirements are addressed. Federal, state, and public participation in decision making are also discussed

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

    Energy Technology Data Exchange (ETDEWEB)

    T. R. Thomas; A. K. Herbst

    2005-06-01

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

  6. Midwestern High-Level Radioactive Waste Transportation Project

    International Nuclear Information System (INIS)

    1993-01-01

    On February 17,1989, the Midwestern Office of The Council of State Governments and the US Department of Energy entered into a cooperative agreement authorizing the initiation of the Midwestern High-Level Radioactive Waste Transportation Project. The transportation project continued to receive funding from DOE through amendments to the original cooperative agreement, with December 31, 1993, marking the end of the initial 5-year period. This progress report reflects the work completed by the Midwestern Office from February 17,1989, through December 31,1993. In accordance with the scopes of work governing the period covered by this report, the Midwestern Office of The Council of State Governments has worked closely with the Midwestern High-Level Radioactive Waste Committee. Project staff have facilitated all eight of the committee's meetings and have represented the committee at meetings of DOE's Transportation Coordination Group (TCG) and Transportation External Coordination Working Group (TEC/WG). Staff have also prepared and submitted comments on DOE activities on behalf of the committee. In addition to working with the committee, project staff have prepared and distributed 20 reports, including some revised reports (see Attachment 1). Staff have also developed a library of reference materials for the benefit of committee members, state officials, and other interested parties. To publicize the library, and to make it more accessible to potential users, project staff have prepared and distributed regular notices of resource availability

  7. Risk communication system for high level radioactive waste disposal

    International Nuclear Information System (INIS)

    Kugo, Akihide; Uda, Akinobu; Shimoda, Hirosi; Yoshikawa, Hidekazu; Ito, Kyoko; Wakabayashi, Yasunaga

    2005-01-01

    In order to gain a better understanding and acceptance of the task of implementing high level radioactive waste disposal, a study on new communication system about social risk information has been initiated by noticing the rapid expansion of Internet in the society. First, text mining method was introduced to identify the core public interest, examining public comments on the technical report of high level radioactive waste disposal. Then we designed the dialog-mode contents based on the theory of norm activation by Schwartz. Finally, the discussion board was mounted on the web site. By constructing such web communication system which includes knowledge base contents, introspective contents, and interactive discussion board, we conducted the experiment for verifying the principles such as that the basic technical knowledge and trust, and social ethics are indispensable in this process to close the perception gap between nuclear specialists and the general public. The participants of the experiment increased their interest in the topics with which they were not familiar and actively posted their opinions on the BBS. The dialog-mode contents were significantly more effective than the knowledge-based contents in promoting introspection that brought people into a greater awareness of problems such as social dilemma. (author)

  8. The principal radionuclides in high level radioactive waste management

    International Nuclear Information System (INIS)

    Mulyanto

    1998-01-01

    The principal radionuclides in high level radioactive waste management. The selection of the principal radionuclides in the high level waste (HLW) management was developed in order to improve the disposal scenario of HLW. In this study the unified criteria for selection of the principal radionuclides were proposed as; (1) the value of hazard index estimated by annual limit of intake (ALI) for long-term tendency,(2) the relative dose factor related to adsorbed migration rate transferred by ground water, and (3) heat generation in the repository. From this study it can be concluded that the principal radionuclides in the HLW management were minor actinide (MA=Np, Am, Cm, etc), Tc, I, Cs and Sr, based on the unified basic criteria introduced in this study. The remaining short-lived fission product (SLFPs), after the selected nuclides are removed, should be immobilized and solidified in a glass matrix. Potential risk due to the remaining SLFPs can be lower than that of uranium ore after about 300 year. (author)

  9. Robotic systems for the high level waste tank farm replacement project at INEL

    International Nuclear Information System (INIS)

    Berger, A.; White, D.; Thompson, B.; Christensen, M.

    1993-01-01

    Westinghouse Idaho Nuclear Company (WINCO) is specifying and designing a new high level waste tank farm at the Idaho National Engineering Laboratory (INEL). The farm consists of four underground storage tanks, which replace the existing tanks. The new facility includes provisions for remote operations. One of the planned remote operations is robotic inspection of the tank from the interior and exterior. This paper describes the process used to design the robotic system for the inspection tasks

  10. Mobile plant for encapsulating of solid high-level radioactive waste in metal matrix

    International Nuclear Information System (INIS)

    Sobolev, I.A.; Arustamov, A.Eh.; Shiryaev, V.V.; Ozhovan, M.I.; Semenov, K.N.; Kachalov, M.B.

    1993-01-01

    Technology for disposal of spent radionuclide sources of ionizing radiation into the standard well-type storage facilities is considered. Universal mobile facility, providing for incorporation of high-level solid wastes into metallic matrices, is proposed. The facility consists of separate moduli, assembled on a transport platform. Electrical meter, wherein the matrix metal (lead and its alloys) is melted and heated up to 600-800 C constitutes the basic modulus in the facility. 4 refs., 4 figs

  11. Thermal analysis of a ventilated high-level waste repository

    International Nuclear Information System (INIS)

    1977-04-01

    The cooling response of a single ventilated storage room in an unventilated array of rooms is examined. Calculations show that ventilation provides a thermal sink in the heated system inducing temperature gradients in the formation different from the unventilated case. An asymptotic cool-down limit exists for the storage room temperature; this minimum temperature depends on inlet air temperature, ventilation flow rate, and convective heat transfer coefficient. For inlet air at 75 0 F and 50,000 cfm and a heat transfer coefficient of 0.8 Btu h- 0 F-ft 2 , the limit is about 100 0 F. A storage room sealed for 5 years will achieve temperatures of approximately 180 0 F, and approximately 4 months would be required in order to cool the storage room floor to a temperature of 120 0 F with a flow rate of 50,000 cfm at an inlet air temperature of 75 0 F, assuming a convective heat transfer coefficient of 0.8 Btu/h- 0 F-ft 2 . Two months would be needed to cool the exhaust air to 120 0 F. For large air flow rates, the cooling time is independent of the flow rate. Increasing the storage room surface area by 25% over the baseline model depresses the cool-down temperatures by only 4 0 F and decreases cooling times by 20%. Modifications in canister design or width have virtually no effect on the cooling, but placing the waste deeper beneath the storage rooms and/or using longer canisters can lower the operating temperatures and cooling times. Reducing the canisters from 3.5 kW power density for 10-year-old waste (108.5 kW/acre) to 2.0 kW/canister (62 kW/acre) reduces cooling temperatures by more than 20 0 F and reduces cooling times to a few weeks or less. The cooling times are nearly independent of the conductivity of the geologic formation. The temperature increase in the air brought from the surface down the supply shaft to the storage room level is about 5 to 7 F 0 per 1000 feet. Temperature increases in regionsshould not be seriously restricted 30 or more feet away

  12. Modeling by GASP-IV simulation of high-level nuclear waste disposal

    International Nuclear Information System (INIS)

    Kurstedt, H.A. Jr.; DePorter, E.L.; Turek, J.L.; Funk, S.K.; Rasbach, C.E.

    1981-01-01

    High-level nuclear waste generated by defense-oriented and commercial nuclear energy activities are to be stored ultimately in underground repositories. Research continues on the waste-form and waste-form processing. DOE managers must coordinate the results of this research, the capacities and availability times of the permanent geologic storage repositories, and the capacities and availability times of interim storage facilities (pending availability of permanent repositories). Comprehensive and active DOE program-management information systems contain predicted generation of nuclear wastes from defense and commercial activities; milestones on research on waste-forms; and milestones on research and development, design, acquisition, and construction of facilities and repositories. A GASP IV simulation model is presented which interfaces all of these data. The model accepts alternate management decisions; relates all critical milestones, all research and development data, and the generation of waste nuclear materials; simulates the passage of time; then, predicts the impact of those alternate decisions on the availability of storage capacity for waste nuclear materials. 3 references, 3 figures

  13. Recovery of Cs from high level radioactive waste

    International Nuclear Information System (INIS)

    Kumar, Amar; Kaushik, C.P.; Raj, K.; Varshney, Lalit

    2008-01-01

    Separation of Cs + from HLW restricts the personal radiation exposure during the vitrification and prevents thermal deformation of conditioned waste matrix during storage because of the high calorific power of 134 Cs (13.18 W/g) and 137 Cs (0.417 W/g) which would markedly reduce the storage cost. Separation will also reduce its volatility during vitrification and extent of migration from the vitrified mass in repository. In addition 137 Cs has enormous applications as radiation sources in food preservation, sterilization of medical products, brachy therapy, blood irradiation, hygienization of sewage sludge etc. The use of 137 Cs (T 1/2 = 30 years) in place of 60 Co (T 1/2 = 5.2 years) will also reduce the shielding requirement and frequency of source replenishment which will ease the handling/transportation of radioactive source

  14. Procedure for conditioning high-level solidified wastes

    Energy Technology Data Exchange (ETDEWEB)

    Hild, W; Krause, H; Scheffler, K

    1974-05-30

    The molds of glass, ceramic or basalt-similar mass in which highly radioactive wastes are incorporated are used for the conditioning of waste waters and/or of sewage or precipitating sludge or of natural water to obtain drinking water, prior to the end storage. By means of the gamma-radiation they emit, the viruses and bacteria and worm eggs are killed off as well as the poisonous, and organic substances such as, e.g., chlorated aromatics are destroyed. Furthermore, the filtration power is increased by coagulation, and the sludge is drained. Natural water is degermed. In particular, fission product mixtures of light water reactors can be incorporated in the molds. The molds are immersed in the media.

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

    Energy Technology Data Exchange (ETDEWEB)

    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).

  16. High level waste transport and disposal cost calculations for the United Kingdom

    International Nuclear Information System (INIS)

    Nattress, P.C.; Ward, R.D.

    1992-01-01

    Commercial nuclear power has been generated in the United Kingdom since 1962, and throughout that time fuel has been reprocessed giving rise to high level waste. This has been managed by storing fission products and related wastes as highly active liquor, and more recently by a program of vitrification and storage of the glass blocks produced. Government policy is that vitrified high level waste should be stored for at least 50 years, which has the technical advantage of allowing the heat output rate of the waste to fall, making disposal easier and cheaper. Thus, there is no immediate requirement to develop a deep geological repository in the UK, but the nuclear companies do have a requirement to make financial provision out of current revenues for high level waste disposal at a future repository. In 1991 the interested organizations undertook a new calculation of costs for such provisions, which is described here. The preliminary work for the calculation included the assumption of host geology characteristics, a compatible repository concept including overpacking, and a range of possible nuclear programs. These have differing numbers of power plants, and differing mixes of high level waste from reprocessing and spent fuel for direct disposal. An algorithm was then developed so that the cost of high level waste disposal could be calculated for any required case within a stated envelope of parameters. An Example Case was then considered in detail leading to the conclusion that a repository to meet the needs of a constant UK nuclear economy up to the middle of the next century would have a cash cost of UK Pounds 1194M (US$2011M). By simple division the cost to a kWh of electricity is UK Pounds 0.00027 (0.45 US mil). (author)

  17. High-level waste management research and development program at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Blomeke, J.O.; Bond, W.D.

    1976-01-01

    Projections of wastes to be generated through the year 2000 portend a problem of impressive size and complexity but one which can be handled within the framework of current and planned investigative programs. Investigations of the technical feasibility of removing actinide elements from wastes to render the residuals more manageable in terms of hazards and storage requirements indicate that they can be removed from wastes by the minimally desired factors of 10 2 to 10 4 ; however, demonstrations and engineering assessments of chemical flowsheets have yet to be made. Natural salt formations are believed to offer the best prospects for disposal of high-level wastes; other promising geological formations are also being evaluated for their suitability for use in the disposal of wastes

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-08-31

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

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

    International Nuclear Information System (INIS)

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

    1980-10-01

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

  20. High-level radioactive waste disposal: Key geochemical issues and information needs for site characterization

    International Nuclear Information System (INIS)

    Brooks, D.J.; Bembia, P.J.; Bradbury, J.W.; Jackson, K.C.; Kelly, W.R.; Kovach, L.A.; Mo, T.; Tesoriero, J.A.

    1986-01-01

    Geochemistry plays a key role in determining the potential of a high-level radioactive waste disposal site for long-term radionuclide containment and isolation. The Nuclear Regulatory Commission (NRC) has developed a set of issues and information needs important for characterizing geochemistry at the potential sites being investigated by the Department of Energy Basalt Waste Isolation Project, Nevada Nuclear Waste Storage Investigations project, and Salt Repository Project. The NRC site issues and information needs consider (1) the geochemical environment of the repository, (2) changes to the initial geochemical environment caused by construction and waste emplacement, and (3) interactions that affect the transport of waste radionuclides to the accessible environment. The development of these issues and information needs supports the ongoing effort of the NRC to identify and address areas of geochemical data uncertainty during prelicensing interactions

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

    International Nuclear Information System (INIS)

    Mendel, J.E.

    1984-08-01

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

  2. Synroc - a multiphase ceramic for high level nuclear waste immobilisation

    International Nuclear Information System (INIS)

    Reeve, K.D.; Vance, E.R.; Hart, K.P.; Smith, K.L.; Lumpkin, G.R.; Mercer, D.J.

    1992-01-01

    Many natural minerals - particularly titanates - are very durable geochemically, having survived for millions of years with very little alteration. Moreover, some of these minerals have quantitatively retained radioactive elements and their daughter products over this time. The Synroc concept mimics nature by providing an all-titanate synthetic mineral phase assemblage to immobilise high level waste (HLW) from nuclear fuel reprocessing operations for safe geological disposal. In principle, many chemically hazardous inorganic wastes arising from industry could also be immobilised in highly durable ceramics and disposed of geologically, but in practice the cost structure of most industries is such that lower cost waste management solutions - for example, the development of reusable by-products or the use of cements rather than ceramics - have to be devised. In many thousands of aqueous leach tests at ANSTO, mostly at 70-90 deg C, Synroc has been shown to be exceptionally durable. The emphases of the recent ANSTO program have been on tailoring of the Synroc composition to varying HLW compositions, leach testing of Synroc containing radioactive transuranic actinides, study of leaching mechanisms by SEM and TEM, and the development and costing of a conceptual fully active Synroc fabrication plant design. A summary of recent results on these topics will be presented. 29 refs., 4 figs

  3. ATW system impact on high-level waste

    International Nuclear Information System (INIS)

    Arthur, E.D.

    1992-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mendel, J.E. (compiler)

    1984-08-01

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

  5. Management of commercial high-level nuclear waste in the United States

    International Nuclear Information System (INIS)

    Mann, S.A.

    1981-01-01

    The Nuclear Waste Policy Act of 1982 (Act), enacted by the 97th Congress in December and signed into law on Jan 7, 1983, by President Reagan, brings a whole new perspective to the nation's nuclear waste management effort. An elaborate set of near-term requirements and actions have to be accomplished within the first 180 days of this Act. As an initial step, Secretary of Energy Donald P. Hodel has established a Nuclear Waste Policy Act (NWPA) project office. The director of the NWPA project office, Robert L. Morgan, is responsible for the department's initial activities to implement the Act until the Office of Civilian Radioactive Waste Management, established by Section 304 of the Act, is activated. The Act requires major efforts in two primary areas: disposal and storage of spent fuel and high-level waste. Efforts in the former area are to be financed by fees collected from utilities and placed into a Nuclear Waste Fund for disposal services. The Act provides for federal storage of up to 1900 tons of spent fuel for those utilities that the Nuclear Regulatory Commission (NRC) determines cannot reasonably provide sufficient additional on-site storage. This federal storage is to be financed through utility-user fees that are placed into an Interim Storage Fund. The Act also provides for cooperative research, development, and demonstration activities at utility sites and federal sites. These activities are to be jointly funded by the utilities and the federal government. Lastly, there are generic research and development (RandD) activities in the spent fuel area that are funded from general appropriations. Mandated milestones have been established by the Act in the areas of gologic repository, fund management, monitored retrievable storage, and spent fuel storage

  6. Storage of radioactive waste

    International Nuclear Information System (INIS)

    Pittman, F.K.

    1974-01-01

    Four methods for managing radioactive waste in order to protect man from its potential hazards include: transmutation to convert radioisotopes in waste to stable isotopes; disposal in space; geological disposal; and surface storage in shielded, cooled, and monitored containers. A comparison of these methods shows geologic disposal in stable formations beneath landmasses appears to be the most feasible with today's technology. (U.S.)

  7. Storage facility for radioactive wastes

    International Nuclear Information System (INIS)

    Okada, Kyo

    1998-01-01

    Canisters containing high level radioactive wastes are sealed in overpacks in a receiving building constructed on the ground. A plurality of storage pits are formed in a layered manner vertically in multi-stages in deep underground just beneath the receiving building, for example underground of about 1000m from the ground surface. Each of the storage pits is in communication with a shaft which vertically communicates the receiving building and the storage pits, and is extended plainly in a horizontal direction from the shaft. The storage pit comprises an overpack receiving chamber, a main gallery and a plurality of galleries. A plurality of holes for burying the overpacks are formed on the bottom of the galleries in the longitudinal direction of the galleries. A plurality of overpack-positioning devices which run in the main gallery and the galleries by remote operation are disposed in the main gallery and the galleries. (I.N.)

  8. Silica based gel as a potential waste form for high level waste from fuel reprocessing

    International Nuclear Information System (INIS)

    Ford, C.E.; Dempster, T.J.; Melling, P.J.

    1983-10-01

    To assess the feasibility of safe disposal of high-level radioactive waste as synthetic clay, or material that would react with ground water to form clay, experiments have been carried out to determine the hydrothermal crystallisation and leaching behaviour of silica based gels fired at 900 deg C. Crystallisation rates at a pressure of 500 bars and at temperatures below 400 deg C are negligible and this more or less precludes pre-disposal production of synthetic clay on the scale required. Leaching experiments suggest that the leach rates of Cs from gels by distilled water are higher than those of boro-silicate glasses and SYNROC at the lower temperatures that would be preferred for geological storage. However, amounts of bulk dissolution of gels may be lower than those of boro-silicate glasses. The initial leaching behaviour of gels might be considerably improved by hot compaction at 900 to 1000 deg C. Consideration of likely waste form dissolution behaviour in a repository environment suggests that gels of appropriate composition might perform as well as, or better than, boro-silicate glasses. A novel hypothetical plant is described that could produce the gel waste form on the scale required on a more or less continuous basis. (author)

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

    International Nuclear Information System (INIS)

    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

  10. Fluidized-bed calcination of simulated commercial high-level radioactive wastes

    International Nuclear Information System (INIS)

    Freeby, W.A.

    1975-11-01

    Work is in progress at the Idaho Chemical Processing Plant to verify process flowsheets for converting simulated commercial high-level liquid wastes to granular solids using the fluidized-bed calcination process. Primary emphasis in the series of runs reported was to define flowsheets for calcining simulated Allied-General Nuclear Services (AGNS) waste and to evaluate product properties significant to calcination, solids storage, or post treatment. Pilot-plant studies using simulated high-level acid wastes representative of those to be produced by Nuclear Fuel Services, Inc. (NFS) are also included. Combined AGNS high-level and intermediate-level waste (0.26 M Na in blend) was successfully calcined when powdered iron was added (to result in a Na/Fe mole ratio of 1.0) to the feed to prevent particle agglomeration due to sodium nitrate. Long-term runs (approximately 100 hours) showed that calcination of the combined waste is practical. Concentrated AGNS waste containing sodium at concentrations less than 0.2 M were calcined successfully; concentrated waste containing 1.13 M Na calcined successfully when powdered iron was added to the feed to suppress sodium nitrate formation. Calcination of dilute AGNS waste by conventional fluid-bed techniques was unsuccessful due to the inability to control bed particle size--both particle size and bed level decreased. Fluid-bed solidification of AGNS dilute waste at conditions in which most of the calcined solids left the calciner vessel with the off-gas was successful. In such a concept, the steady-state composition of the bed material would be approximately 22 wt percent calcined solids deposited on inert particles. Calcination of simulated NFS acid waste indicated that solidification by the fluid-bed process is feasible

  11. Problems related to final disposal of high-level radioactive waste in Russia

    International Nuclear Information System (INIS)

    Velichkin, Vasily I.

    1999-01-01

    According to this presentation, the radioactivity of the total amount of radioactive waste accumulated in Russia to date is 1.5*10 9 Ci and of spent fuel 4.5*10 9 Ci. A table is given that shows the source, type, volume activity and storage type under the responsibility of the different departments and enterprises. 99.9% of the wastes are accumulated at the enterprises of Minatom of the Russian Federation. Some companies inject their liquid wastes from ionisation sources and intermediate liquid waste from the nuclear power industry into deep-seated reliably isolated aquifers. The Mayak plant has released liquid low-level and intermediate wastes into artificial reservoirs and Lake Karachay. Liquid high-level wastes are always stored in special tanks at interim storage facilities. A large number of nuclear submarines are laid up in North-Western Russia and East Russia, with spent fuel still in place as the interim storages in these regions are filled up and there are no conditioning plants. Underground disposal is considered the best way of isolating radioactive waste for as long as it is hazardous to the environment. Two new technologies are discussed. One involves including long-lived isotopes in high-stable mineral matrices, the other uses selective separation from the bulk of wastes. The matrices should be disposed of deep in the Earth's crust, at least 2-3 km down. Liquid waste of caesium-strontium fraction must be transformed into glass-like form and stored underground at a depth of a few hundred metres. Short-lived low level and intermediate level wastes should be conditioned and then deposited in subsurface ferroconcrete repositories constructed in clays. Finally, the presentation discusses the selection of sites and conditions for radioactive waste disposal. Two sites are discussed, the Mayak plant and a possible site at Mining Chemical Combine in Krasnoyarsk-26

  12. Phase behavior and radiation effects in high level waste class

    International Nuclear Information System (INIS)

    Turcotte, R.P.; Roberts, F.P.

    1977-02-01

    Results are presented that demonstrate that detailed and reproducible data can be obtained for complex waste glasses. For the major glass composition examined, thermal treatment was shown to cause formation of several crystalline phases which contribute to an increased leachability. Although not discussed in detail here, Zn 2 SiO 4 formation results in microcracking due to a thermal expansion mismatch with the glass matrix, and SrMoO 4 has a higher leachability than the glass matrix. The temperature dependence describing equilibrium concentrations of these two phases and a qualitative understanding of ingrowth kinetics have been established, hence conditions necessary to eliminate their formation during processing and early storage, are known. Radiation damage effects, when extrapolated to long times, suggest energy storage of approximately 50 cal/gram and either positive or negative density changes occur (depending on the glass composition) in the 1 percent range. No radiation damage-related changes of serious concern have been found for homogeneous glasses by 244 Cm doping experiments now approaching a simulated damage time of approximately 10 3 years (for UO 2 fuel wastes). More work is needed concerning heterogeneous damage which will occur in devitrified glasses. As a final point, the complications with respect to understanding behavior of polyphase systems with respect to either radiation damage or leaching behavior, are self evident. Homogeneous glasses with improved leach resistance, and thermal and radiation stability are clear objectives for future glass development

  13. Magnox waste storage complex

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    This article looks at the design and construction of British Nuclear Fuel Limited's (BNFL) Magnox waste storage complex by Costain Engineering Limited. Magnox swarf from fuel decanning is stored underwater in specially designed silos. Gas processing capabilities from Costain Engineering Limited and the experience of BNFL combined in this project to provide the necessary problem-solving skills necessary for this waste storage upgrading and extension project. A retrofitted inerting facility was fitted to an existing building and a new storage extension was fitted, both without interrupting reprocessing operations at Sellafield. (UK)

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  15. Indian program on management of high level radioactive waste - emphasis on value recovery for societal applications

    International Nuclear Information System (INIS)

    Kaushik, C.P.; Tomar, Neelima Singh; Kumar, Amar; Wadhwa, S.; Diwan, Jyoti

    2017-01-01

    Nuclear Power Programme in India is based on 'closed fuel cycle'. Closed fuel cycle involves reprocessing and recycling of Spent Nuclear Fuel (SNF) coming out of nuclear reactors. During reprocessing, uranium and plutonium, constituting bulk of the SNF are separated and subsequently recycled. The remaining small portion constitutes high level radioactive waste containing most of the fission products and minor actinides. A three-step strategy involving immobilization, interim storage followed by ultimate disposal has been adopted in India for management of High Level Waste (HLW). Borosilicate glass matrix has been identified for immobilization of HLW owing to optimal waste loading, adequate leach resistance and long term stability of the product. An interim storage facility is in operation for storage and surveillance of VWP. A comprehensive program based on screening of different materials like granite, argillite, clay with respect to sorption of different radionuclides is being pursued to identify the suitable areas of disposal of the conditioned waste products. Separation of useful radionuclides like "1"3"7Cs, "9"0Sr, "9"0Y, "1"0"6Ru etc and its utilization for societal applications is being practiced in India. (author)

  16. Hypothetical accidents at disposal facilities for high-level liquid radioactive wastes and pulps

    International Nuclear Information System (INIS)

    Kabakchi, S.A.; Zagainov, V.A.; Lishnikov, A.A.; Nazin, E.R.

    1994-01-01

    Four accidents are postulated and analyzed for interim storage of high-level, liquid radioactive wastes at a fuel reprocessing facility. Normal waste storage operation is based on wastes stored in steel drums, partially buried in concrete canyons, and equipped with heat exchangers for cooling and ventilation systems for removal of explosive gases and vapors. The accident scenarios analyzed are: (1) shutdown of ventilation with open entrance and exit ventilation pipes, (2) shutdown of ventilation with closed entrance and exit ventilation pipes, (3) shutdown of the cooling system with normally functioning ventilation, and (4) simultaneous cooling and ventilation system failure (worst case). A mathematical model was developed and used to calculate radiation consequences of various accidents. Results are briefly presented for the worst case scenario and compared to an actual accident for model validation. 17 refs., 3 figs., 1 tab

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  18. Partitioning high-level waste from alkaline solution: A literature survey

    International Nuclear Information System (INIS)

    Marsh, S.F.

    1993-05-01

    Most chemical partitioning procedures are designed for acidic feed solutions. However, the high-level waste solutions in the underground storage tanks at US Department of Energy defense production sites are alkaline. Effective partitioning procedures for alkaline solutions could decrease the need to acidify these solutions and to dissolve the solids in acid, which would simplify subsequent processing and decrease the generation of secondary waste. The author compiles candidate technologies from his review of the chemical literature, experience, and personal contacts. Several of these are recommended for evaluation

  19. High-level Waste Long-term management technology development

    International Nuclear Information System (INIS)

    Choi, Jong Won; Kang, C. H.; Ko, Y. K.

    2012-02-01

    The purpose of this project is to develop a long-term management system(A-KRS) which deals with spent fuels from domestic nuclear power stations, HLW from advanced fuel cycle and other wastes that are not admitted to LILW disposal site. Also, this project demonstrate the feasibility and reliability of the key technologies applied in the A-KRS by evaluating them under in-situ condition such as underground research laboratory and provide important information to establish the safety assessment and long-term management plan. To develop the technologies for the high level radioactive wastes disposal, demonstrate their reliability under in-situ condition and establish safety assessment of disposal system, The major objects of this project are the following: Ο An advanced disposal system including waste containers for HLW from advanced fuel cycle and pyroprocess has been developed. Ο Quantitative assessment tools for long-term safety and performance assessment of a radwaste disposal system has been developed. Ο Hydrological and geochemical investigation and interpretation methods has been developed to evaluate deep geological environments. Ο The THMC characteristics of the engineered barrier system and near-field has been evaluated by in-situ experiments. Ο The migration and retardation of radionuclides and colloid materials in a deep geological environment has been investigated. The results from this project will provide important information to show HLW disposal plan safe and reliable. The knowledge from this project can also contribute to environmental conservation by applying them to the field of oil and gas industries to store their wastes safe

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    Weber, A.H.

    1999-01-01

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

  2. Thermal characteristics of rocks for high-level waste repository

    International Nuclear Information System (INIS)

    Shimooka, Kenji; Ishizaki, Kanjiro; Okamoto, Masamichi; Kumata, Masahiro; Araki, Kunio; Amano, Hiroshi

    1980-12-01

    Heat released by the radioactive decay of high-level waste in an underground repository causes a long term thermal disturbance in the surrounding rock mass. Several rocks constituting geological formations in Japan were gathered and specific heat, thermal conductivity, thermal expansion coefficient and compressive strength were measured. Thermal analysis and chemical analysis were also carried out. It was found that volcanic rocks, i.e. Andesite and Basalt had the most favorable thermal characteristics up to around 1000 0 C and plutonic rock, i.e. Granite had also favorable characteristics under 573 0 C, transition temperature of quartz. Other igneous rocks, i.e. Rhyolite and Propylite had a problem of decomposition at around 500 0 C. Sedimentary rocks, i.e. Zeolite, Tuff, Sandstone and Diatomite were less favorable because of their decomposition, low thermal conductivity and large thermal expansion coefficient. (author)

  3. Transmutation of high-level radioactive waste - Perspectives

    CERN Document Server

    Junghans, Arnd; Grosse, Eckart; Hannaske, Roland; Kögler, Toni; Massarczyk, Ralf; Schwengner, Ronald; Wagner, Andreas

    2014-01-01

    In a fast neutron spectrum essentially all long-lived actinides (e.g. Plutonium) undergo fission and thus can be transmuted into generally short lived fission products. Innovative nuclear reactor concepts e.g. accelerator driven systems (ADS) are currently in development that foresee a closed fuel cycle. The majority of the fissile nuclides (uranium, plutonium) shall be used for power generation and only fission products will be put into final disposal that needs to last for a historical time scale of only 1000 years. For the transmutation of high-level radioactive waste a lot of research and development is still required. One aspect is the precise knowledge of nuclear data for reactions with fast neutrons. Nuclear reactions relevant for transmutation are being investigated in the framework of the european project ERINDA. First results from the new neutron time-of-flight facility nELBE at Helmholtz-Zentrum Dresden-Rossendorf will be presented.

  4. High-level radioactive-waste-disposal investigations in Texas

    International Nuclear Information System (INIS)

    Smith, R.D.

    1983-01-01

    The Texas Energy and Natural Resources Advisory Council (TENRAC) was designated in 1980 to coordinate the interaction between the State of Texas and the federal government relating to the high-level radioactive waste disposal issue. This report was prepared to summarize the many aspects of that issue with particular emphasis on the activities in Texas. The report is intended to provide a comprehensive introduction for individuals with little or no previous exposure to the issue and to provide a broader perspective for those individuals who have addressed specific aspects of the issue but have not had the opportunity to study it in a broader context. Following the introduction, contents of this report are as follows: (1) general status of major repository siting investigations in the US; (2) detailed review of Texas studies; (3) possible facilities to be sited in Texas; (4) current Texas policy; (5) federal regulations; and (6) federal legislation. 9 figures, 2 tables

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

    International Nuclear Information System (INIS)

    Gannon, R.E.

    1993-01-01

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

  6. Geology of high-level nuclear waste disposal

    International Nuclear Information System (INIS)

    Roxburgh, I.S.

    1988-01-01

    The concept of geological disposal is set out by describing the major rock types in terms of their ability to isolate high-level nuclear waste. The advantages and problems posed by particular rock formations are explored and the design and construction of geological repositories is considered, along with the methods used to estimate their safety. It gives special consideration to the use of sea-covered rock and sediment as well as the on-land situation. Throughout the book the various principles and problems inherent in geological disposal are explained and illustrated by reference to a multitude of European and North American case studies, backed up by a large number of tables, figures and an extensive bibliography

  7. Risk assessment methodology for Hanford high-level waste tanks

    International Nuclear Information System (INIS)

    Bott, T.F.; Mac Farlane, D.R.; Stack, D.W.; Kindinger, J.

    1992-01-01

    A methodology is presented for applying Probabilistic Safety Assessment techniques to quantification of the health risks posed by the high-level waste (HLW) underground tanks at the Department of Energy's Hanford reservation. This methodology includes hazard screening development of a list of potential accident initiators, systems fault trees development and quantification, definition of source terms for various release categories, and estimation of health consequences from the releases. Both airborne and liquid pathway releases to the environment, arising from aerosol and spill/leak releases from the tanks, are included in the release categories. The proposed methodology is intended to be applied to a representative subset of the total of 177 tanks, thereby providing a baseline risk profile for the HLW tank farm that can be used for setting clean-up/remediation priorities. Some preliminary results are presented for Tank 101-SY

  8. Midwestern High-Level Radioactive Waste Transportation Project

    International Nuclear Information System (INIS)

    Sattler, L.R.

    1992-02-01

    In addition to arranging for storage and disposal of radioactive waste, the US Department of Energy (DOE) must develop a safe and efficient transportation system in order to deliver the material that has accumulated at various sites throughout the country. The ability to transport radioactive waste safely has been demonstrated during the past 20 years: DOE has made over 2,000 shipments of spent fuel and other wastes without any fatalities or environmental damage related to the radioactive nature of the cargo. To guarantee the efficiency of the transportation system, DOE must determine the optimal combination of rail transport (which allows greater payloads but requires special facilities) and truck transport Utilizing trucks, in turn, calls for decisions as to when to use legal weight trucks or, if feasible, overweight trucks for fewer but larger shipments. As part of the transportation system, the Facility Interface Capability Assessment (FICA) study contributes to DOE's development of transportation plans for specific facilities. This study evaluates the ability of different facilities to receive, load and ship the special casks in which radioactive materials will be housed during transport In addition, the DOE's Near-Site Transportation Infrastructure (NSTI) study (forthcoming) will evaluate the rail, road and barge access to 76 reactor sites from which DOE is obligated to begin accepting spent fuel in 1998. The NSTI study will also assess the existing capabilities of each transportation mode and route, including the potential for upgrade

  9. Management of commercial high-level and transuranium-contaminated radioactive wastes. Environmental statement

    International Nuclear Information System (INIS)

    1974-09-01

    This Draft Environmental Statement is issued to assess the environmental impact of the AEC's program to manage commercial high-level and transuranium-contaminated radioactive wastes. These are the types of commercial radioactive wastes for which AEC custody is required by present or anticipated regulations. The program consists of three basic parts: development of a Retrievable Surface Storage Facility (RSSF) for commercial high-level waste, using existing technology; evaluating geological formations and sites for the development of a Geological Disposal Pilot Plant (GDPP) which would lead to permanent disposal; and providing retrievable storage for the transuranium-contaminated waste pending availability of permanent disposal. Consideration has been given to all environmental aspects of the program, using waste generation projections through the year 2000. Radiological and other impacts of implementing the program are expected to be minimal, but will be discussed in further environmental statements which will support budget actions for specific repositories. The alternatives discussed in this Draft Environmental Statement are presented. (U.S.)

  10. Vitrification of low level and mixed (radioactive and hazardous) wastes: Lessons learned from high level waste vitrification

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1994-01-01

    Borosilicate glasses will be used in the USA and in Europe immobilize radioactive high level liquid wastes (HLLW) for ultimate geologic disposal. Simultaneously, tehnologies are being developed by the US Department of Energy's (DOE) Nuclear Facility sites to immobilize low-level and mixed (radioactive and hazardous) wastes (LLMW) in durable glass formulations for permanent disposal or long-term storage. Vitrification of LLMW achieves large volume reductions (86--97 %) which minimize the associated long-term storage costs. Vitrification of LLMW also ensures that mixed wastes are stabilized to the highest level reasonably possible, e.g. equivalent to HLLW, in order to meet both current and future regulatory waste disposal specifications The tehnologies being developed for vitrification of LLMW rely heavily on the technologies developed for HLLW and the lessons learned about process and product control

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    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

  13. Underground excavation methods for a high-level waste repository

    International Nuclear Information System (INIS)

    Peshel, J.; Gupta, D.; Nataraja, M.

    1990-01-01

    This paper reports on rock excavation methods for a High-Level Waste repository that should be selected to limit the potential for creating preferential pathways for groundwater to travel to the waste packages or for radionuclides to migrate to the accessible environment. The use of water and other foreign substances should be controlled so that the repository performance is not compromised. The excavated openings should remain stable so that operations can be carried out safely and the retrievability option maintained. As per the current conceptual designs presented by the Department of Energy, the exploratory shaft facility becomes a part of the repository if the Yucca Mountain site is found suitable for repository development. Therefore, the methods of constructing the underground openings should be compatible with the performance requirements for the repository. Also, the degree of damage to the rock surrounding the openings and the extent of the damage zone should not preclude adequate site characterization. The ESf construction and operation should be compatible with the site data gathering activities, such as geological, thermomechanical, hydrological and geochemical testing

  14. The disposal of high-level radioactive waste. Vol. 1

    International Nuclear Information System (INIS)

    Parker, F.L.; Broshears, R.E.; Pasztor, J.

    1984-01-01

    The Beijer Institute received request from the Swedish Board for Spent Nuclear Fuel (Naemnden for Anvaent Kaernbraensle - NAK) to undertake an international review of the major programmes which were currently making arrangements for the future disposal of high-level radioactive wastes and spent nuclear fuel. The request was accepted, a detailed proposal was worked out and agreed to by NAK, for a critical technical review which concentrated on the following three main tasks: 1. a 'state-of-the-art' review of selected ongoing disposal programmes, both national and international; 2. an assessment of the scientific and technical controversies involved, and 3. recommendations for further research in this field. This review work was to be built on a survey of the available technical literature which was to serve as a basis for a series of detailed interviews, consultations and discussions with scientific and technical experts in Japan, Canada, USA, Belgium, Federal Republic of Germany, France, Switzerland and the United Kingdom. This first volume contains: disposal options; review of the state-of-the-art (international activities, national programs); analysis of waste disposal systems. (orig./HP)

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

    International Nuclear Information System (INIS)

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

    1996-03-01

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

  16. Safety of geological disposal of high-level waste

    International Nuclear Information System (INIS)

    Ohe, Toshiaki; Tsukamoto, Masaki

    1989-01-01

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

  17. The IFR pyroprocessing for high-level waste minimization

    International Nuclear Information System (INIS)

    Laidler, J.J.

    1993-01-01

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

  18. Control of high-level radioactive waste-glass melters

    International Nuclear Information System (INIS)

    Bickford, D.F.; Coleman, C.J.

    1990-01-01

    The Defense Waste Processing Facility (DWPF) will immobilize Savannah River Site High Level Waste as a durable borosilicate glass for permanent disposal in a repository. The DWPF will be controlled based on glass composition. The following discussion is a preliminary analysis of the capability of the laboratory methods that can be used to control the glass composition, and the relationships between glass durability and glass properties important to glass melting. The glass durability and processing properties will be controlled by controlling the chemical composition of the glass. The glass composition will be controlled by control of the melter feed transferred from the Slurry Mix Evaporator (SME) to the Melter Feed Tank (MFT). During cold runs, tests will be conducted to demonstrate the chemical equivalence of glass sampled from the pour stream and glass removed from cooled canisters. In similar tests, the compositions of glass produced from slurries sampled from the SME and MFT will be compared to final product glass to determine the statistical relationships between melter feed and glass product. The total error is the combination of those associated with homogeneity in the SME or MFT, sampling, preparation of samples for analysis, instrument calibration, analysis, and the composition/property model. This study investigated the sensitivity of estimation of property data to the combination of variations from sampling through analysis. In this or a similar manner, the need for routine glass product sampling will be minimized, and glass product characteristics will be assured before the melter feed is committed to the melter

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Immobilization of high-level wastes into sintered glass: 2

    International Nuclear Information System (INIS)

    Bevilacqua, A.M.; Russo, D.O.; Messi de Bernasconi, N.; Audero, M.A.

    1987-01-01

    High level radioactive wastes are immobilized into borosilicate glasses. Experiences with the variety VG 98/12 (SiO 2 , TiO 2 , Al 2 O 3 , B 2 O 3 , MgO, CaO, Na 2 O) are described. The pressing was performed in a matrix of 12.7 mm diameter, the walls of which were lubricated with sterotex dissolved in Cl 4 C. The sintering was made in an horizontal electric furnace in air atmosphere at temperatures between 500 and 600 deg C. It was observed that the maximum density occurs at 605 deg C. Comparing both the hot and the cold pressing process, it is concluded that: 1) In spite of requiring more complex equipment the hot pressing process has the advantage that lower pressures are applied, with the consequent obtainment of waste blocks with larger diameters, and 2) it is advisable that pressing process should be performed in the definitive can. (M.E.L.) [es

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  2. Mathematical modelling of heat production in deep geological repository of high-level nuclear waste

    International Nuclear Information System (INIS)

    Kovanda, O.

    2017-01-01

    Waste produced by nuclear industry requires special handling. Currently, there is a research taking place, focused at possibilities of nuclear waste storage in deep geological repositories, hosted in stable geological environment. The high-level nuclear waste produces significant amount of heat for a long time, which can affect either environment outside of or within the repository in a negative way. Therefore to reduce risks, it is desirable to know the principles of such heat production, which can be achieved using mathematical modeling. This thesis comes up with a general model of heat production-time dependency, dependable on initial composition of the waste. To be able to model real situations, output of this thesis needs to be utilized in an IT solution. (authors)

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

    International Nuclear Information System (INIS)

    Cash, R.J.

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    Valenti, P.J.; Elliott, D.I.

    1999-01-01

    This report provides a summary of operational experiences, component and system performance, and lessons learned associated with the operation of the Vitrification Facility (VF) at the West Valley Demonstration Project (WVDP). The VF was designed to convert stored high-level radioactive waste (HLW) into a stable waste form (borosilicate glass) suitable for disposal in a federal repository. Following successful completion on nonradioactive test, HLW processing began in July 1995. Completion of Phase 1 of HLW processing was reached on 10 June 1998 and represented the processing of 9.32 million curies of cesium-137 (Cs-137) and strontium-90 (Sr-90) to fill 211 canisters with over 436,000 kilograms of glass. With approximately 85% of the total estimated curie content removed from underground waste storage tanks during Phase 1, subsequent operations will focus on removal of tank heel wastes

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

    Energy Technology Data Exchange (ETDEWEB)

    WILLIS, W.L.

    2000-06-15

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

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

    International Nuclear Information System (INIS)

    WILLIS, W.L.

    2000-01-01

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

  7. Final repositories for high-level radioactive waste; Endlagerung hochradioaktiver Abfaelle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2015-10-15

    The brochure on final repositories for high-level radioactive waste covers the following issues: What is the origin of radioactive wastes? How large are the waste amounts? What is going to happen with the wastes? What is the solution for the Waste disposal? A new site search is started. Safety requirements for the final disposal of high-level radioactive wastes. Comparison of host rocks. Who is responsible and who will pay? Final disposal of high-level radioactive wastes worldwide. Short summary: History of the search for a final repository for high-level radioactive wastes in Germany.

  8. Transportation of high-level waste and spent fuel

    International Nuclear Information System (INIS)

    Carlson, J.H.; Lake, W.H.; Thompson, J.H.

    1993-01-01

    The Office of Civilian Radioactive Waste Management (OCRWM) transportation program is a multifaceted undertaking to transport spent nuclear fuel from commercial reactors to temporary and permanent storage facilities commencing in 1998. One of the significant ingredients necessary to achieving this goal is the development and acquisition of shipping casks. Efforts to design and acquire high capacity casks is ongoing, as are efforts to purchase casks that can be made available using current technology. By designing casks that are optimized to the specifications of the older cooler spent fuel that will be shipped, and by designing to current NRC requirements, OCRWM's new generation of spent fuel casks will be more efficient and at least as safe as current cask designs. (J.P.N.)

  9. Assessment of studies and researches on warehousing - High-level and intermediate-level-long-lived radioactive wastes - December 2012

    International Nuclear Information System (INIS)

    2013-01-01

    This large report first presents the approach adopted for the study and research on the warehousing of high-level and intermediate-level-long-lived radioactive wastes. It outlines how reversible storage and warehousing are complementary, discusses the lessons learned from researches performed by the CEA on long duration warehousing, presents the framework of studies and researches performed since 2006, and presents the scientific and technical content of studies and researches (warehousing need analysis, search for technical options providing complementarity with storage, extension or creation of warehousing installations). The second part addresses high-level and intermediate-level-long-lived radioactive waste parcels, indicates their origins and quantities. The third part proposes an analysis of warehousing capacities: existing capacities, French industrial experience in waste parcel warehousing, foreign experience in waste warehousing. The fourth part addresses reversible storage in deep geological formation: storage safety functions, storage reversibility, storage parcels, storage architecture, chronicle draft. The fifth part proposes an inventory of warehousing needs in terms of additional capacities for the both types of wastes (high-level, and intermediate-level-long-lived), and discusses warehousing functionalities and safety objectives. The sixth and seventh parts propose a detailed overview of design options for warehousing installations, respectively for high-level and for intermediate-level-long-lived waste parcels: main technical issues, feasibility studies of different concepts or architecture shapes, results of previous studies and introduction to studies performed since 2011, possible evolutions of the HA1, HA2 and MAVL concepts. The eighth chapter reports a phenomenological analysis of warehousing and the optimisation of material selection and construction arrangements. The last part discusses the application of researches to the extension of the

  10. Description of a ceramic waste form and canister for Savannah River Plant high-level waste

    International Nuclear Information System (INIS)

    Butler, J.L.; Allender, J.S.; Gould, T.H. Jr.

    1982-04-01

    A canistered ceramic waste form for possible immobilization of Savannah River Plant (SRP) high-level radioactive wastes is described. Characteristics reported for the form include waste loading, chemical composition, heat content, isotope inventory, mechanical and thermal properties, and leach rates. A conceptual design of a potential production process for making this canistered form are also described. The ceramic form was selected in November 1981 as the primary alternative to the reference waste form, borosilicate glass, for making a final waste form decision for SRP waste by FY-1983. 11 tables

  11. The importance of international cooperation in the field of high level radioactive waste management

    International Nuclear Information System (INIS)

    Isaacs, Thomas H.

    1992-01-01

    This paper discusses the importance of international collaboration in the field of radioactive waste management and points out how cooperation has benefited the U.S. civilian waste management program. The U.S. Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) oversees the handling, transportation, storage, and final deposition of high-level radioactive wastes for the U.S. commercial sector. Because OCRWM shares many of the same waste management concerns as various other countries with nuclear programs, and since one country's waste management program will ultimately have an impact on the waste management programs of other countries, it is clearly in the interest of all countries to work together in search of solutions to common waste management problems. To facilitate this. cooperation, OCRWM is a participating member of international organizations, such as the IAEA and the OECD/NEA. OCRWM further has in place several bilateral agreements with various individual countries and with the Commission of the European Communities (CEC). Other international waste management initiatives are also currently being considered. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-10-12

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

    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

  14. Standard format and content for a license application to store spent fuel and high-level radioactive waste

    International Nuclear Information System (INIS)

    1989-09-01

    Subpart B, ''License Application, Form, and Contents,'' of 10 CFR Part 72, ''Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste,'' specifies the information to be covered in an application for a license to store spent fuel in an independent spent fuel storage installation (ISFSI) or to store spent fuel and high-level radioactive waste in a monitored retrievable storage facility (MRS). However, Part 72 does not specify the format to be followed in the license application. This regulatory guide suggests a format acceptable to the NRC staff for submitting the information specified in Part 72 for license application to store spent fuel in an ISFSI or to store spent fuel and high-level radioactive waste in an MRS

  15. Corrosion issues in high-level nuclear waste containers

    Science.gov (United States)

    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

  16. Mineralogic studies of tuff for high-level waste disposal

    International Nuclear Information System (INIS)

    Vaniman, D.; Bish, D.; Broxton, D.; Byers, F.; Carlos, B.; Levy, S.

    1986-01-01

    The volcanic rocks at Yucca Mountain, Nevada, consist predominantly of tuff that originated 12 to 14 million years ago as flows and airfalls of hot volcanic particulates. On cooling these units formed two major rock types: crystallized zones formed mostly of feldspar and silica minerals, and zones of glass. Alteration of glass to zeolite minerals occurred largely during structural tilting of Yucca Mountain in the ∼1-3 million years following the major eruptions. The compositions of zeolites formed from glasses strongly indicate open-system chemical exchange. Superimposed on this general alteration of glasses are areas of local high-temperature alteration. High-temperature alteration ended by 11 million years ago. Zeolites such as clinoptilolite persisted during high-temperature alteration at temperatures up to 100 degree C, suggesting that clinoptilolite at Yucca Mountain close to the thermally disturbed zone around a repository may also survive heating to temperatures at least this high. The mineralogic data from tuff at Yucca Mountain will ultimately be used by the Department of Energy Nevada Nuclear Waste storage Investigations for (1) defining the mineralogic component in estimating waste element travel times away from the repository and (2) determining the past history of alteration and the anticipated stability of minerals near the repository

  17. United States high-level radioactive waste management program: Current status and plans

    International Nuclear Information System (INIS)

    Williams, J.

    1992-01-01

    The inventory of spent fuel in storage at reactor sites in the United States is approximately 20,000 metric tons heavy metal (MTHM). It is increasing at a rate of 1700 to 2100 MTHM per year. According to current projections, by the time the last license for the current generation of nuclear reactors expires, there will be an estimated total of 84,000 MTHm. No commercial reprocessing capacity exists or is planned in the US. Therefore, the continued storage of spent fuel is required. The majority of spent fuel remains in the spent fuel pools of the utilities that generated it. Three utilities are presently supplementing pool capacity with on-site dry storage technologies, and four others are planning dry storage. Commercial utilities are responsible for managing their spent fuel until the Federal waste management system, now under development, accepts spent fuel for storage and disposal. Federal legislation charges the Office of Civilian Radioactive Waste Management (OCRWM) within the US Department of Energy (DOE) with responsibility for developing a system to permanently dispose of spent fuel and high level radioactive waste in a manner that protects the health and safety of the public and the quality of the environment. We are developing a waste management system consisting for three components: a mined geologic repository, with a projected start date of 2010; a monitored retrievable storage facility (MRS), scheduled to begin waste acceptance in 1998; and a transportation system to support MRS and repository operations. This paper discusses the background and framework for the program, as well as the current status and plans for management of spent nuclear fuel at commercial utilities; the OCRWM's development of a permanent geologic repository, an MRS, and a transportation system; the OCRWM's safety approach; the OCRWM's program management initiatives; and the OCRWM's external relations activities

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

    International Nuclear Information System (INIS)

    Cornman, W.R.

    1981-12-01

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

  19. Study on the development of safety regulations for geological disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    Wei Fangxin

    2012-01-01

    The development of regulations under Regulations on Safety Management of Radioactive Waste has become necessary as the issuance of it. The regulations related to geological disposal of high-level radioactive waste can promote the progress of research and development on geological disposal of high-level radioactive waste in China. This paper has present suggestions on development of regulations on geological disposal of high-level radioactive waste by analyzing development of safety regulations on geological disposal of high-level radioactive waste in foreign countries and problems occurred in China and discussed important issues related to the development of safety regulations on geological disposal of high-level radioactive waste. (author)

  20. Thermophysical properties of Hanford high-level tank wastes: A preliminary survey of recent data

    International Nuclear Information System (INIS)

    Willingham, C.E.

    1994-03-01

    This report documents an analysis performed by Pacific Northwest Laboratory (PNL) involving thermophysical properties of Hanford high-level tank wastes. PNL has gathered and summarized the available information on density, viscosity, thermal conductivity, heat capacity, particle size, shear strength, and heat generation. The information was compiled from documented characterization reports of Hanford single-shell and double-shell tanks. The report summarizes the thermophysical properties of the various waste materials, the anticipated range for the various waste forms, and estimates of the variability of the measured data. The thermophysical information compiled in this study is useful as input to sensitivity and parametric studies for the Multi-Function Waste Tank Facility Project. Information from only 33 of the 177 high-level waste storage tanks was compiled. Density data are well characterized for the tanks selected in this study. It was found that the reported viscosity of the wastes varies widely and that a single value should not be used to represent viscosity for all waste. Significant variations in reported shear strength and heat generation values were also found. Very few of the tank characterization reports described information on waste heat capacity. In addition, there was no supernatant vapor pressure information reported in the waste characterization reports examined in this study. Although thermal conductivity measurements were made for a number of tanks, most of the measurements were made in 1975. Finally, particle size distribution measurements of waste in 20 tanks were compiled. The analyst must be cognizant of differences between the number and volume distributions reported for particle size

  1. Final disposal of high-level radioactive waste. State of knowledge and development for safety assessment

    International Nuclear Information System (INIS)

    Sato, Seichi; Muraoka, Susumu; Murano, Toru

    1995-01-01

    In Europe and USA, the formation disposal of high level radioactive waste entered the stage of doing the activities aiming at its execution. Also in Japan, the storage of high level waste began in the spring of 1995. Regarding the utilization of nuclear power, the establishment of the technology for disposing radioactive waste is the subject of fist priority, and the stage that requires its social recognition has set in. There are the features of formation disposal in that the disposal is in the state of confining extremely large amount of radioactivity, and that the assessment of long term safety exceeding tens of thousands years is demanded. The amount of occurrence and the main nuclides of high level radioactive waste, the disposal as seen in the Coady report and in the IAEA standard, the selection of dispersion or confinement and the selection of passive system or long term human participation, the reason why formation disposal is selected, the features of formation disposal and the way of advancing the research, the general techniques of safety assessment, artificial barriers and natural barriers for formation disposal, and the subjects of formation disposal are described. (K.I.) 57 refs

  2. High-level waste issues and resolutions document

    International Nuclear Information System (INIS)

    1994-05-01

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

  3. The precautionary principle and high-level nuclear waste policy

    International Nuclear Information System (INIS)

    Frishman, S.

    1999-01-01

    The 'Precautionary Principle' has grown from the broadening observation that there is compelling evidence that damage to humans and the world-wide environment is of such a magnitude and seriousness that new principles for conducting human activities are necessary. One of the various statements of the Precautionary Principle is: when an activity raises threats of harm to human health or the environment, precautionary measures should be taken even if some cause and effect relationships are not fully established scientifically. The use of a precautionary principle was a significant recommendation emerging from the 1992 United Nations Conference on Environment and Development, held in Rio de Janeiro, Brazil, and it is gaining acceptance in discussions ranging from global warming to activities that affect the marine environment, and far beyond. In the US high-level nuclear waste policy, there is a growing trend on the part of geologic repository proponents and regulators to shift the required safety evaluation from a deterministic analysis of natural and engineered barriers and their interactions to risk assessments and total system waste containment and isolation performance assessment. This is largely a result of the realisation that scientific 'proof' of safety cannot be demonstrated to the level repository proponents have led the American public to expect. Therefore, they are now developing other methods in an attempt to effectively lower the repository safety expectations of the public. Implicit in this shift in demonstration of 'proof' is that levels of uncertainty far larger than those generally taken as scientifically acceptable must be accepted in repository safety, simply because greater certainty is either too costly, in time and money, or impossible to achieve at the potential Yucca Mountain repository site. In the context of the Precautionary Principle, the repository proponent must bear the burden of providing 'Acceptable' proof, established by an open

  4. Characterization and reaction behavior of ferrocyanide simulants and Hanford Site high-level ferrocyanide waste

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Simpson, B.C.

    1994-02-01

    Nonradioactive waste simulants and initial ferrocyanide tank waste samples were characterized to assess potential safety concerns associated with ferrocyanide high-level radioactive waste stored at the Hanford Site in underground single-shell tanks (SSTs). Chemical, physical, thermodynamic, and reaction properties of the waste simulants were determined and compared to properties of initial samples of actual ferrocyanide wastes presently in the tanks. The simulants were shown to not support propagating reactions when subjected to a strong ignition source. The simulant with the greatest ferrocyanide concentration was shown to not support a propagating reaction that would involve surrounding waste because of its high water content. Evaluation of dried simulants indicated a concentration limit of about 14 wt% disodium mononickel ferrocyanide, below which propagating reactions could not occur in the ambient temperature bulk tank waste. For postulated localized hot spots where dried waste is postulated to be at an initial temperature of 130 C, a concentration limit of about 13 wt% disodium mononickel ferrocyanide was determined, below which propagating reactions could not occur. Analyses of initial samples of the presently stored ferrocyanide waste indicate that the waste tank ferrocyanide concentrations are considerably lower than the limit for propagation for dry waste and that the water content is near that of the as-prepared simulants. If the initial trend continues, it will be possible to show that runaway ferrocyanide reactions are not possible under present tank conditions. The lower ferrocyanide concentrations in actual tank waste may be due to tank waste mixing and/or degradation from radiolysis and/or hydrolysis, which may have occurred over approximately 35 years of storage

  5. Radioactive waste storage and disposal: the challenge

    International Nuclear Information System (INIS)

    Prince, A.T.

    1978-03-01

    Solutions to waste management problems are available. After radium is removed, tailings from uranium ores can be disposed of safely in well-designed retention areas. Work is being done on the processing of non-fuel reactor wastes through incineration, reverse osmosis, and evaporation. Spent fuels have been stored safely for years in pools; dry storage in concrete cannisters is being investigated. Ultimate disposal of high-level wastes will be in deep, stable geologic formations. (LL)

  6. Statement of work for the immobilized high-level waste transportation system, Project W-464

    Energy Technology Data Exchange (ETDEWEB)

    Mouette, P.

    1998-06-24

    The objective of this Statement of Work (SOW) is to present the scope, the deliverables, the organization, the technical and schedule expectations for the development of a Package Design Criteria (PDC), cost and schedule estimate for the acquisition of a transportation system for the Immobilized High-Level Waste (IHLW). This transportation system which includes the truck, the trailer, and a shielded cask will be used for on-site transportation of the IHLW canisters from the private vendor vitrification facility to the Hanford Site interim storage facility, i.e., vaults 2 and 3 of the Canister Storage Building (CSB). This Statement of Work asks Waste Management Federal Services, Inc., Northwest Operations, to provide Project W-464 with a Design Criteria Document, plus a life-cycle schedule and cost estimate for the acquisition of a transportation system (shielded cask, truck, trailer) for IHLW on-site transportation.

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

    International Nuclear Information System (INIS)

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

    1994-01-01

    The disposition of actinide elements released from high-level waste glasses into a tuff groundwater in laboratory tests at 90 degrees C at various glass surface area/leachant volume ratios (S/V) between dissolved, suspended, and sorbed fractions has been measured. While the maximum release of actinides is controlled by the corrosion rate of the glass matrix, their solubility and sorption behavior affects the amounts present in potentially mobile phases. Actinide solubilities are affected by the solution pH and the presence of complexants released from the glass, such as sulfate, phosphate, and chloride, radiolytic products, such as nitrate and nitrite, and carbonate. Sorption onto inorganic colloids formed during lass corrosion may increase the amounts of actinides in solution, although subsequent sedimentation of these colloids under static conditions leads to a significant reduction in the amount of actinides in solution. The solution chemistry and observed actinide behavior depend on the S/V of the test. Tests at high S/V lead to higher pH values, greater complexant concentrations, and generate colloids more quickly than tests at low S/V. The S/V also affects the rate of glass corrosion

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

    International Nuclear Information System (INIS)

    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

  9. Site suitability criteria for solidified high level waste repositories

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  10. Concentration of High Level Radioactive Liquid Waste. Basic data acquisition

    Energy Technology Data Exchange (ETDEWEB)

    Juvenelle, A.; Masson, M.; Garrido, M.H. [DEN/VRH/DRCP/SCPS/LPCP, BP 17171 - 30207 Bagnols sur Ceze Cedex (France)

    2008-07-01

    Full text of publication follows: In order to enhance its knowledge about the concentration of high level liquid waste (HLLW) from the nuclear fuel reprocessing process, a program of studies was defined by Cea. In a large field of acidity, it proposes to characterize the concentrated solution and the obtained precipitates versus the concentration factor. Four steps are considered: quantification of the salting-out effect on the concentrate acidity, acquisition of solubility data, precipitates characterisation versus the concentration factor through aging tests and concentration experimentation starting from simulated fission products solutions. The first results, reported here, connect the acidity of the concentrated solution to the concentration factor and allow us to precise the field of acidity (4 to 12 N) for the next experiments. In this field, solubility data of various elements (Ba, Sr, Zr...) are separately measured at room temperature, in nitric acid in a first time, then in the presence of various species present in medium (TBP, PO{sub 4}{sup 3-}). The reactions between these various elements are then investigated (formation of insoluble mixed compounds) by following the concentration cations in solution and characterising the precipitates. (authors)

  11. Why consider subseabed disposal of high-level nuclear wastes

    International Nuclear Information System (INIS)

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

    1983-01-01

    There exist large areas of the deep seabed that warrant assessment as potential disposal sites for high-level radioactive wastes because (1) they are far from seismically and tectonically active lithospheric plate boundaries; (2) they are far from active or young volcanoes; (3) they contain thick layers of very uniform fine-grained clays; (4) they are devoid of natural resources likely to be exploited in the foreseeable future; (5) the geologic and oceanographic processes governing the deposition of sediments in such areas are well understood, and have been remarkably insensitive to past oceanic and climatic changes; and (6) sedimentary 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 clayey sediments indicate that they can act as a primary barrier to the escape of buried radionuclides. 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

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

    International Nuclear Information System (INIS)

    Bell, Jimmy T.

    2000-01-01

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

  13. Alternatives to high-level waste vitrification: The need for common sense

    International Nuclear Information System (INIS)

    Bell, J.T.

    2000-01-01

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

  14. Characterization and vitrification of Hanford radioactive high level wastes

    International Nuclear Information System (INIS)

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

    1991-01-01

    Radioactive Neutralized Current Acid Waste (NCAW) samples from the Hanford waste tanks have been chemically, radiochemically and physically characterized. The wastes were processed according to the Hanford Waste vitrification Plant (HWVP) flowsheet, and characterized after each process step. The waste glasses were sectioned and leach tested. Chemical, radiochemical and physical properties of the waste will be presented and compared to nonradioactive simulant data and the HWVP reference composition and properties

  15. Waste package designs for disposal of high-level waste in salt formations

    International Nuclear Information System (INIS)

    Basham, S.J. Jr.; Carr, J.A.

    1984-01-01

    In the United States of America the selected method for disposal of radioactive waste is mined repositories located in suitable geohydrological settings. Currently four types of host rocks are under consideration: tuff, basalt, crystalline rock and salt. Development of waste package designs for incorporation in mined salt repositories is discussed. The three pertinent high-level waste forms are: spent fuel, as disassembled and close-packed fuel pins in a mild steel canister; commercial high-level waste (CHLW), as borosilicate glass in stainless-steel canisters; defence high-level waste (DHLW), as borosilicate glass in stainless-steel canisters. The canisters are production and handling items only. They have no planned long-term isolation function. Each waste form requires a different approach in package design. However, the general geometry and the materials of the three designs are identical. The selected waste package design is an overpack of low carbon steel with a welded closure. This container surrounds the waste forms. Studies to better define brine quantity and composition, radiation effects on the salt and brines, long-term corrosion behaviour of the low carbon steel, and the leaching behaviour of the spent fuel and borosilicate glass waste forms are continuing. (author)

  16. 324 Radiochemical engineering cells and high level vault tanks mixed waste compliance status

    International Nuclear Information System (INIS)

    1994-01-01

    The 324 Building in the Hanford 300 Area contains Radiochemical Engineering Cells and High Level Vault tanks (the open-quotes REC/HLVclose quotes) for research and development activities involving radioactive materials. Radioactive mixed waste within this research installation, found primarily in B-Cell and three of the high level vault tanks, is subject to RCRA/DWR (open-quotes RCRAclose quotes) regulations for storage. This white paper provides a baseline RCRA compliance summary of MW management in the REC/HLV, based on best available knowledge. The REC/HLV compliance project, of which this paper is a part, is intended to achieve the highest degree of compliance practicable given the special technical difficulties of managing high activity radioactive materials, and to assure protection of human health and safety and the environment. The REC/HLV was constructed in 1965 to strict standards for the safe management of highly radioactive materials. Mixed waste in the REC/HLV consists of discarded tools and equipment, dried feed stock from nuclear waste melting experiments, contaminated particulate matter, and liquid feed stock from various experimental programs in the vault tanks. B-Cell contains most of these materials. Total radiological inventory in B-Cell is estimated at 3 MCi, about half of which is potentially open-quotes dispersibleclose quotes, that is, it is in small pieces or mobile particles. Most of the mixed waste currently in the REC/HLV was generated or introduced before mixed wastes were subjected to RCRA in 1987

  17. High-level nuclear waste borosilicate glass: A compendium of characteristics

    International Nuclear Information System (INIS)

    Cunnane, J.C.; Bates, J.K.; Ebert, W.L.; Feng, X.; Mazer, J.J.; Wronkiewicz, D.J.; Sproull, J.; Bourcier, W.L.; McGrail, B.P.

    1992-01-01

    With the imminent startup, in the United States, of facilities for vitrification of high-level nuclear waste, a document has been prepared that compiles the scientific basis for understanding the alteration of the waste glass products under the range of service conditions to which they may be exposed during storage, transportation, and eventual geologic disposal. A summary of selected parts of the content of this document is provided. Waste glass alterations in a geologic repository may include corrosion of the glass network due to groundwater and/or water vapor contact. Experimental testing results are described and interpreted in terms of the underlying chemical reactions and physical processes involved. The status of mechanistic modeling, which can be used for long-term predictions, is described and the remaining uncertainties associated with long-term simulations are summarized

  18. Radiation doses in alternative commercial high-level waste management systems

    International Nuclear Information System (INIS)

    Schneider, K.J.; Pelto, P.J.; Lavender, J.C.; Daling, P.M.; Fecht, B.A.

    1986-01-01

    In the commercial high-level waste management system, potential changes are being considered that will augment the benefits of an integral monitored retrievable storage (MRS) facility. The US Department of Energy (DOE) has recognized that alternative options could be implemented in the authorized waste management system (i.e., without an integral MRS facility) to potentially achieve some of the same beneficial effects of the integral MRS system. This paper summarizes those DOE-sponsored analyses related to radiation doses resulting from changes in the waste management system. This report presents generic analyses of aggregated radiation dose impacts to the public and occupational workers, of nine postulated changes in the operation of a spent-fuel management system without an MRS facility

  19. Numerical modeling of heat outflux from a vitrified high level waste

    International Nuclear Information System (INIS)

    Aravind, Arun; Jayaraj, Aparna; Seshadri, H.; Balasubramaniyan, V.

    2018-01-01

    Heat generating vitrified high-level waste is initially stored in interim storage facility with adequate cooling for sufficient period of time, and then proposed to be disposed of in deep geological repositories. Heat flux from the waste form can cause thermo mechanical changes within the disposal module and also in the surrounding rock. It may change the permeability of rock fractures over a period of time. It is very essential to study the long term performance of deep geological repository to build confidence in the design and over all operation of the disposal facility. In this study a numerical model was developed to study the temperature distribution in the waste matrix and also the heat out flux to the surrounding rock matrix

  20. Incorporation of high-level nuclear waste in gel spheres

    International Nuclear Information System (INIS)

    Robinson, S.M.; Arnold, W.D.; Bond, W.D.; Angelini, P.; Stinton, D.P.

    1981-01-01

    Waste sludge is incorporated in gel spheres by the method of internal gelation. Gel spheres containing up to 90 wt % waste have been produced from defense and commercial wastes. A generic cesium-bearing waste form has been developed. Pyrolytic carbon and SiC coatings reduce the leachability of all tested articles to the detection limits

  1. Actinides and fission products partitioning from high level liquid waste

    International Nuclear Information System (INIS)

    Yamaura, Mitiko

    1999-01-01

    The presence of small amount of mixed actinides and long-lived heat generators fission products as 137 Cs and 90 Sr are the major problems for safety handling and disposal of high level nuclear wastes. In this work, actinides and fission products partitioning process, as an alternative process for waste treatment is proposed. First of all, ammonium phosphotungstate (PWA), a selective inorganic exchanger for cesium separation was chosen and a new procedure for synthesizing PWA into the organic resin was developed. An strong anionic resin loaded with tungstate or phosphotungstate anion enables the precipitation of PWA directly in the resinous structure by adding the ammonium nitrate in acid medium (R-PWA). Parameters as W/P ratio, pH, reactants, temperature and aging were studied. The R-PWA obtained by using phosphotungstate solution prepared with W/P=9.6, 9 hours digestion time at 94-106 deg C and 4 to 5 months aging time showed the best capacity for cesium retention. On the other hand, Sr separation was performed by technique of extraction chromatography, using DH18C6 impregnated on XAD7 resin as stationary phase. Sr is selectively extracted from acid solution and >99% was recovered from loaded column using distilled water as eluent. Concerning to actinides separations, two extraction chromatographic columns were used. In the first one, TBP(XAD7) column, U and Pu were extracted and its separations were carried-out using HNO 3 and hydroxylamine nitrate + HNO 3 as eluent. In the second one, CMP0-TBP(XAD7) column, the actinides were retained on the column and the separations were done by using (NH 4 ) 2 C 2 O 4 , DTPA, HNO 3 and HCl as eluent. The behavior of some fission products were also verified in both columns. Based on the obtained data, actinides and fission products Cs and Sr partitioning process, using TBP(XAD7) and CMP0-TBP(XAD7) columns for actinides separation, R-PWA column for cesium retention and DH18C6(XAD7) column for Sr isolation was performed

  2. Embedding of solid high-level wastes into metal and non-metal matrices

    International Nuclear Information System (INIS)

    Geel, J. van; Eschrich, H.; Dobbels, F.; Favre, P.; Sterner, H.

    1980-03-01

    The primary objective of embedding solidification high-level waste forms of high specific activity into a matrix material is to obtain final waste composites with moderate inner temperatures, even at large waste loadings per meter cylinder length. Secondary objectives are to produce a non-porous, crack-free composite product with a durability superior to that of the embedded waste form itself. The temperature distribution in composite material composed of vitreous beads embedded into a metal matrix (vitromets) are compared with that in a vitreous block, of equal heat generation per meter height, during short- and long-term storage. It was found that for storage under water, inner temperatures below 100 0 C are assured in vitromets, produced from short-cooled high-level wastes, and containing high waste loadings per metercanister height. The chemical and mechanical stability, as well as the thermal conductivity have been examined for vitromets containing various matrix materials whereby emphasis is imparted to lead- and aluminum alloys. The corrosion of lead- and aluminum alloys in distilled water, brine solution and dry salt has been examined at temperatures up to 230 0 C and pressures up to 3.5 MPa. Some lead alloys were found to exhibit superior corrosion resistance in these chemical environments than certain reference borosilicate glasses. The deformation behavior of vitromets under axial compression has been investigated at different temperatures and varying height diameter ratios. The maturity of the vitromet production is finally demonstrated by presenting process data from hot-laboratory scale and cold semi-industrial scale production units. (author)

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

    International Nuclear Information System (INIS)

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

    1996-10-01

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

  4. High-level radioactive waste disposal in the deep ocean

    International Nuclear Information System (INIS)

    Hill, H.W.

    1977-01-01

    A joint programme has begun between the Fisheries Laboratory, Lowestoft and the Institute of Oceanographic Sciences, Wormley to study the dispersion of radioactivity in the deep ocean arising from the possible dumping of high level waste on the sea bed in vitrified-glass form which would permit slow leakage over a long term scale. The programme consists firstly of the development of a simple diffusion/advection model for the dispersion of radioactivity in a closed and finite ocean, which overcomes many of the criticisms of the earlier model proposed by Webb and Morley. Preliminary results from this new model are comparable to those of the Webb-Morley model for radio isotopes with half-lives of 10-300 years but are considerably more restrictive outside this range, particularly for those which are much longer-lived. The second part of the programme, towards which the emphasis is directed, concerns the field programme planned to measure the advection and diffusion parameters in the deeper layers of the ocean to provide realistic input parameters to the model and increase our fundamental understanding of the environment in which the radioactive materials may be released. The first cruises of the programme will take place in late 1976 and involve deep current meter deployments and float dispersion experiments around the present NEA dump site with some sediment sampling, so that adsorption experiments can be started on typical deep sea sediments. The programme will expand the number of long-term deep moored stations over the next five years and include further float experiments, CTD profiling, and other physical oceanography. In the second half of the 5-year programme, attempts will be made to measure diffusion parameters in the deeper layers of the ocean using radioactive tracers

  5. Electrochemical probing of high-level radioactive waste tanks containing washed sludge and precipitates

    International Nuclear Information System (INIS)

    Bickford, D.F.; Congdon, J.W.; Oblath, S.B.

    1986-12-01

    At the US Department of Energy's Savannah River Plant, corrosion of carbon steel storage tanks containing alkaline, high-level radioactive waste is controlled by specification of limits on waste composition and temperature. Processes for the preparation of waste for final disposal will result in waste with low corrosion inhibitor concentrations and, in some cases, high aromatic organic concentrations, neither of which are characteristic of previous operations. Laboratory tests, conducted to determine minimum corrosion inhibitor levels indicated pitting of carbon steel near the waterline for proposed storage conditions. In situ electrochemical measurements of full-scale radioactive process demonstrations have been conducted to assess the validity of laboratory tests. Probes included pH, Eh (potential relative to a standard hydrogen electrode), tank potential, and alloy coupons. In situ results are compared to those of the laboratory tests, with particular regard given to simulated solution composition. Transition metal hydroxide sludge contains strong passivating species for carbon steel. Washed precipitate contains organic species that lower solution pH and tend to reduce passivating films, requiring higher inhibitor concentrations than the 0.01 molar nitrite required for reactor fuel reprocessing wastes. Periodic agitation, to keep the organic phase suspended, or cathodic protection are possible alternatives to higher nitrite inhibitor concentrations

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  7. Interaction of cementitious materials with high-level waste

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

    1978-05-01

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

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

    International Nuclear Information System (INIS)

    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

  10. Comparison of SRP high-level waste disposal costs for borosilicate glass and crystalline ceramic waste forms

    International Nuclear Information System (INIS)

    McDonell, W.R.

    1982-04-01

    An evaluation of costs for the immobilization and repository disposal of SRP high-level wastes indicates that the borosilicate glass waste form is less costly than the crystalline ceramic waste form. The wastes were assumed immobilized as glass with 28% waste loading in 10,300 reference 24-in.-diameter canisters or as crystalline ceramic with 65% waste loading in either 3400 24-in.-diameter canisters or 5900 18-in.-diameter canisters. After an interim period of onsite storage, the canisters would be transported to the federal repository for burial. Total costs in undiscounted 1981 dollars of the waste disposal operations, excluding salt processing for which costs are not yet well defined, were about $2500 million for the borosilicate glass form in reference 24-in.-diameter canisters, compared to about $2900 million for the crystalline ceramic form in 24-in.-diameter canisters and about $3100 million for the crystalline ceramic form in 18-in.-diameter canisters. No large differences in salt processing costs for the borosilicate glass and crystalline ceramic forms are expected. Discounting to present values, because of a projected 2-year delay in startup of the DWPF for the crystalline ceramic form, preserved the overall cost advantage of the borosilicate glass form. The waste immobilization operations for the glass form were much less costly than for the crystalline ceramic form. The waste disposal operations, in contrast, were less costly for the crystalline ceramic form, due to fewer canisters requiring disposal; however, this advantage was not sufficient to offset the higher development and processing costs of the crystalline ceramic form. Changes in proposed Nuclear Regulatory Commission regulations to permit lower cost repository packages for defense high-level wastes would decrease the waste disposal costs of the more numerous borosilicate glass forms relative to the crystalline ceramic forms

  11. Project Guarantee 1985. Final repository for high-level radioactive wastes: The system of safety barriers

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    Final disposal of radioactive waste involves preventing the waste from returning from the repository location into the biosphere by means of successively arranged containment measures known as safety barriers. In the present volume NGB 85-04 of the series of reports for Project 'Guarantee' 1985, the safety barrier system for the type C repository for high-level waste is described. The barrier parameters which are relevant for safety analysis are quantified and associated error limits and data scatter are given. The aim of the report is to give a summary documentation of the safety analysis input data and their scientific background. For secure containment of radioactive waste safety barriers are used which effectively limit the release of radioactive material from the repository (release barriers) and effectively retard the entry of the original radioactive material into the biosphere (time barriers). Safety barriers take the form of both technically constructed containment measures and the siting of the repository in suitable geological formations. The technical safety barrier system in the case of high-level waste comprises: the waste solidification matrix (borosilicate glass), massive steel canisters, encasement of the waste canisters, encasement of the waste canisters in highly compacted bentonite, sealing of vacant storage space and access routes on repository closure. The natural geological safety barriers - the host rock and overlying formations provide sufficiently long deep groundwater flow times from the repository location to the earth's surface and for additional lengthening of radionuclide migration times by means of various chemical and physical retardation mechanisms. The stability of the geological formations is so great that hydrogeological system is protected for a sufficient length of time from deterioration caused, in particular, by erosion. Observations in the final section of the report indicate that input data for the type C repository safety

  12. Seismic scoping evaluation of high level liquid waste tank vaults at the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Hashimoto, P.S.; Uldrich, E.D.; McGee, W.D.

    1991-01-01

    A seismic scoping evaluation of buried vaults enclosing high level liquid waste storage tanks at the Idaho Chemical Processing Plant has been performed. The objective of this evaluation was to scope out which of the vaults could be demonstrated to be seismically adequate against the Safe Shutdown Earthquake (SSE). Using approximate analytical methods, earthquake experience data, and engineering judgement, this study determined that one vault configuration would be expected to meet ICPP seismic design criteria, one would not be considered seismically adequate against the SSE, and one could be shown to be seismically adequate against the SSE using nonlinear analysis

  13. Borosilicate glass as a matrix for immobilization of SRP high-level waste

    International Nuclear Information System (INIS)

    Wicks, G.G.

    1980-01-01

    Approximately 22 million gallons of high-level radioactive defense waste are currently being stored in large underground tanks located on the Savannah River Plant (SRP) site in Aiken, South Carolina. One option now being considered for long-term management of this waste involves removing the waste from the tanks, chemically processing the waste, and immobilizing the potentially harmful radionuclides in the waste into a borosilicate glass matrix. The technology for producing waste glass forms is well developed and has been demonstrated on various scales using simulated as well as radioactive SRP waste. Recently, full-scale prototypical equipment has been made operational at SRP. This includes both a joule-heated ceramic melter and an in-can melter. These melters are a part of an integrated vitrification system which is under evaluation and includes a spray calciner, direct liquid feed apparatus, and various elements of an off-gas system. Two of the most important properties of the waste glass are mechanical integrity and leachability. Programs are in progress at SRL aimed at minimizing thermally induced cracking by carefully controlling cooling cycles and using ceramic liners or coatings. The leachability of SRP waste glass has been studied under many different conditions and consistently found to be low. For example, the leachability of actual SRP waste glass was found to be 10 -6 to 10 -5 g/(cm 2 )(day) initially and decreasing to 10 -9 to 10 -8 g/(cm 2 )(day) after 100 days. Waste glass is also being studied under anticipated storage conditions. In brine at 90 0 C, the leachability is about 5 x 10 -8 g/(cm 2 )(day) after 60 days. The effects of other geological media including granite, basalt, shale, and tuff are also being studied as part of the multibarrier isolation system

  14. Technology development for long-term management of Hanford high-level waste. Quarterly report, January 1976--March 1976

    Energy Technology Data Exchange (ETDEWEB)

    Kounts, J S [ed.

    1976-06-01

    This document is the seventh in a series of quarterly reports on work in support of long-term management of Hanford high-level wastes. The work reported here was performed during the period January through March 1976. The specific topics discussed are grouped into the subject areas of: Storage System Integrity and Engineered Improvements; Waste Retrieval; Waste Immobilization and Storage; Contaminated Equipment Volume Reduction; and Analytical Methods Development. Previous reports were published as ARH-ST-110 A through D and ARH-ST-132 A and B.

  15. Technological demonstrators. Researches and studies on the storage and disposal of long living intermediate level and high level radioactive wastes; Les demonstrateurs technologiques. Recherches et etudes sur le stockage et l'entreposage des dechets de haute activite et de moyenne activite a vie longue

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-07-01

    This brochure presents the technological demonstrators made by the French national agency of radioactive wastes (ANDRA) and exhibited at Limay (Yvelines, France). These demonstrators, built at scale 1, have been an essential support to the establishment of the 'Dossier 2005' which demonstrates the feasibility of a reversible disposal of long living-intermediate level and high level radioactive wastes in the Callovo-Oxfordian argillite of Meuse-Haute Marne. Two type of demonstrators were built: demonstrators of storage containers for long living-intermediate level wastes and for spent fuels, and dynamic demonstrators for containers handling. This brochure presents these different demonstrators, their characteristics and the results of their tests. (J.S.)

  16. Disposal of high level radioactive wastes in geological formations

    International Nuclear Information System (INIS)

    Martins, L.A.M.; Carvalho Bastos, J.P. de

    1978-01-01

    The disposal of high-activity radioactive wastes is the most serious problem for the nuclear industry. Among the solutions, the disposal of wastes in approriated geological formations is the most realistic and feasible. In this work the methods used for geological disposal, as well as, the criteria, programs and analysis for selecting a bite for waste disposal are presented [pt

  17. Proposed methods for treating high-level pyrochemical process wastes

    International Nuclear Information System (INIS)

    Johnson, T.R.; Miller, W.E.; Steunenberg, R.K.

    1985-01-01

    This survey illustrates the large variety and number of possible techniques available for treating pyrochemical wastes; there are undoubtedly other process types and many variations. The choice of a suitable process is complicated by the uncertainty as to what will be an acceptable waste form in the future for both TRU and non-TRU wastes

  18. Predicted peak temperature-rises around a high-level radioactive waste canister emplaced in the deep ocean bed

    International Nuclear Information System (INIS)

    Kipp, K.L.

    1978-06-01

    A simple mathematical model of heat conduction was used to evaluate the peak temperature-rise along the wall of a canister of high-level radioactive waste buried in deep ocean sediment. Three different amounts of vitrified waste, corresponding to standard Harvest, large Harvest, and AVM canisters, and three different waste loadings were studied. Peak temperature-rise was computed for the nine cases as a function of canister geometry and storage time between reprocessing and burial. Lower waste loadings or longer storage times than initially envisaged are necessary to prevent the peak temperature-rise from exceeding 200 0 C. The use of longer, thinner cylinders only modestly reduces the storage time for a given peak temperature. Effects of stacking of waste canisters and of close-packing were also studied. (author)

  19. Risk comparison of different treatment and disposal strategies of high level liquid radioactive waste

    International Nuclear Information System (INIS)

    Fang Dong

    1997-01-01

    The risk of different treatment and disposal strategies of high level liquid radioactive waste from spent fuel reprocessing is estimated and compared. The conclusions obtained are that risk difference from these strategies is very small and high level liquid waste can be reduced to middle and low level waste, if the decontamination factor for 99 Tc is large enough, which is the largest risk contributor in the high level radioactive waste from spent fuel reprocessing. It is also shown that the risk of high level radioactive waste could be reduced by the technical strategy of combining partitioning and transmutation

  20. Radioactive Waste Management Research Program Plan for high-level waste: 1987

    International Nuclear Information System (INIS)

    1987-05-01

    This plan will identify and resolve technical and scientific issues involved in the NRC's licensing and regulation of disposal systems intended to isolate high level hazardous radioactive wastes (HLW) from the human environment. The plan describes the program goals, discusses the research approach to be used, lays out peer review procedures, discusses the history and development of the high level radioactive waste problem and the research effort to date and describes study objectives and research programs in the areas of materials and engineering, hydrology and geochemistry, and compliance assessment and modeling. The plan also details the cooperative interactions with international waste management research programs. Proposed Earth Science Seismotectonic Research Program plan for radioactive waste facilities is appended