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Sample records for contaminated waste canisters

  1. Development of a facility for fabricating nuclear waste canisters from radioactively contaminated steel

    International Nuclear Information System (INIS)

    Logan, J.A.; Larsen, M.M.

    1986-01-01

    This paper describes design of a facility and processes capable of using radioactively contaminated waste steel as the principal raw material for fabricating stainless steel canisters to be used for disposal of nuclear high-level waste. By such action, expenditure (i.e., permanent loss to society) of thousands of tons of uncontaminated chromium and nickel to fabricate such canisters can be avoided. Moreover, the cost and risks involved in disposing of large accumulations of radioactively contaminated steel as low-level radioactive waste (LLRW), that would otherwise be necessary, can also be avoided. The canister fabrication processes (involving centrifugal casting) described herein have been tested and proven for this application. The performance characteristics of stainless steel canisters so fabricated have been tested and agreed to by the organizations that have been involved in this development work (Battelle Memorial Institute, DuPont, EGandG and the Savannah River Laboratory) as equivalent to the performance characteristics of canisters fabricated of uncontaminated wrought stainless steel. It is estimated that the production cost for fabricating canisters by the methods described will not differ greatly from the production cost using uncontaminated wrought steel, and the other costs avoided by not having to dispose of the contaminated steel as LLRW could cause this method to produce the lowest ultimate overall costs

  2. DISPOSABLE CANISTER WASTE ACCEPTANCE CRITERIA

    Energy Technology Data Exchange (ETDEWEB)

    R.J. Garrett

    2001-07-30

    The purpose of this calculation is to provide the bases for defining the preclosure limits on radioactive material releases from radioactive waste forms to be received in disposable canisters at the Monitored Geologic Repository (MGR) at Yucca Mountain. Specifically, this calculation will provide the basis for criteria to be included in a forthcoming revision of the Waste Acceptance System Requirements Document (WASRD) that limits releases in terms of non-isotope-specific canister release dose-equivalent source terms. These criteria will be developed for the Department of Energy spent nuclear fuel (DSNF) standard canister, the Multicanister Overpack (MCO), the naval spent fuel canister, the High-Level Waste (HLW) canister, the plutonium can-in-canister, and the large Multipurpose Canister (MPC). The shippers of such canisters will be required to demonstrate that they meet these criteria before the canisters are accepted at the MGR. The Quality Assurance program is applicable to this calculation. The work reported in this document is part of the analysis of DSNF and is performed using procedure AP-3.124, Calculations. The work done for this analysis was evaluated according to procedure QAP-2-0, Control of Activities, which has been superseded by AP-2.21Q, Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities. This evaluation determined that such activities are subject to the requirements of DOE/RW/0333P, Quality Assurance Requirements and Description (DOE 2000). This work is also prepared in accordance with the development plan titled Design Basis Event Analyses on DOE SNF and Plutonium Can-In-Canister Waste Forms (CRWMS M&O 1999a) and Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages (CRWMS M&O 2000d). This calculation contains no electronic data applicable to any electronic data management system.

  3. Decontamination processes for waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1982-01-01

    A Defense Waste Processing Facility (DWPF) is currently being designed to convert Savannah River Plant liquid, high-level radioactive waste into a solid form, such as borosilicate glass. To prevent the spread of radioactivity, the outside of the canisters of waste glass must have very low levels of smearable radioactive contamination before they are removed from the DWPF. Several techniques were considered for canister decontamination: high-pressure water spray, electropolishing, chemical dissolution, and abrasive blasting. An abrasive blasting technique using a glass frit slurry has been selected for use in the DWPF. No additional equipment is needed to process waste generated from decontamination. Frit used as the abrasive will be mixed with the waste and fed to the glass melter. In contrast, chemical and electrochemical techniques require more space in the DWPF, and produce large amounts of contaminated by-products, which are difficult to immobilize by vitrification

  4. Waste canister for storage of nuclear wastes

    Science.gov (United States)

    Duffy, James B.

    1977-01-01

    A waste canister for storage of nuclear wastes in the form of a solidified glass includes fins supported from the center with the tips of the fins spaced away from the wall to conduct heat away from the center without producing unacceptable hot spots in the canister wall.

  5. Waste canister for storage of nuclear wastes

    International Nuclear Information System (INIS)

    Duffy, J.B.

    1977-01-01

    A waste canister for storage of nuclear wastes in the form of a solidified glass includes fins supported from the center with the tips of the fins spaced away from the wall to conduct heat away from the center without producing unacceptable hot spots in the canister wall. 4 claims, 4 figures

  6. Friction welded closures of waste canisters

    International Nuclear Information System (INIS)

    Klein, R.F.

    1987-01-01

    Liquid radioactive waste presently stored in underground tanks is to undergo a vitrifying process which will immobilize it into a solid form. This solid waste will be contained in a stainless steel canister. The canister opening requires a positive-seal weld, the properties and thickness of which must be at least equal to those of the canister material. All studies and tests performed in the work discussed in this paper have the inertia friction welding concept to be highly feasible in this application. This paper describes the decision to investigate the inertia friction welding process, the inertia friction welding process itself, and a proposed equipment design concept. This system would provide a positive, reliable, inspectable, and full-thickness seal weld while utilizing easily maintainable equipment. This high-quality weld can be achieved even in highly contaminated hot cell

  7. Decontamination processes for waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1981-06-01

    The process which will be used to decontaminate waste glass canisters at the Savannah River Plant consists of: decontamination (slurry blasting); rinse (high-pressure water); and spot decontamination (high-pressure water plus slurry). No additional waste will be produced by this process because glass frit used in decontamination will be mixed with the radioactive waste and fed into the glass melter. Decontamination of waste glass canisters with chemical and abrasive blasting techniques was investigated. The ability of a chemical technique with HNO 3 -HF and H 2 C 2 O 4 to remove baked-on contamination was demonstrated. A correlation between oxide removal and decontamination was observed. Oxide removal and, thus, decontamination by abrasive blasting techniques with glass frit as the abrasive was proposed and demonstrated

  8. Decontamination processes for waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1981-01-01

    The process which will be used to decontaminate waste glass canisters at the Savannah River Plant consists of: decontamination (slurry blasting); rinse (high-pressure water); and spot decontamination (high-pressure water plus slurry). No additional waste will be produced by this process because glass frit used in decontamination will be mixed with the radioactive waste and fed into the glass melter. Decontamination of waste glass canisters with chemical and abrasive blasting techniques was investigated. The ability of a chemical technique with HNO 3 -HF and H 2 C 2 O 4 to remove baked-on contamination was demonstrated. A correlation between oxide removal and decontamination was observed. Oxide removal and, thus, decontamination by abrasive blasting techniques with glass frit as the abrasive was proposed and demonstrated

  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. Canister arrangement for storing radioactive waste

    Science.gov (United States)

    Lorenzo, D.K.; Van Cleve, J.E. Jr.

    1980-04-23

    The subject invention relates to a canister arrangement for jointly storing high level radioactive chemical waste and metallic waste resulting from the reprocessing of nuclear reactor fuel elements. A cylindrical steel canister is provided with an elongated centrally disposed billet of the metallic waste and the chemical waste in vitreous form is disposed in the annulus surrounding the billet.

  11. Decontamination of Savannah River Plant waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1982-01-01

    A Defense Waste Processing Facility (DWPF) is currently being designed to convert Savannah River Plant (SRP) liquid, high-level radioactive waste into a solid form, such as borosilicate glass. The outside of the canisters of waste glass must have very low levels of smearable radioactive contamination before they are removed from the DWPF to prevent the spread of radioactivity. Several techniques were considered for canister decontamination: high-pressure water spray, electropolishing, chemical dissolution, and abrasive blasting. An abrasive blasting technique using a glass frit slurry has been selected for use in the DWPF. No additional equipment is needed to process waste generated from decontamination. Frit used as the abrasive will be mixed with the waste and fed to the glass melter. In contrast, chemical and electrochemical techniques require more space in the DWPF, and produce large amounts of contaminated byproducts which are difficult to immobilize by vitrification

  12. Chemical compatibility of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    Harbour, J.R.

    1993-01-01

    The Waste Acceptance Preliminary Specifications (WAPS) require that the contents of the canistered waste form are compatible with one another and the stainless steel canister. The canistered waste form is a closed system comprised of a stainless steel vessel containing waste glass, air, and condensate. This system will experience a radiation field and an elevated temperature due to radionuclide decay. This report discusses possible chemical reactions, radiation interactions, and corrosive reactions within this system both under normal storage conditions and after exposure to temperatures up to the normal glass transition temperature, which for DWPF waste glass will be between 440 and 460 degrees C. Specific conclusions regarding reactions and corrosion are provided. This document is based on the assumption that the period of interim storage prior to packaging at the federal repository may be as long as 50 years

  13. Waste canister closure welding using the inertia friction welding process

    International Nuclear Information System (INIS)

    Klein, R.F.; Siemens, D.H.; Kuruzar, D.L.

    1986-02-01

    Liquid radioactive waste presently stored in underground tanks is to undergo a vitrifying process which will immobilize it in a solid form. This solid waste will be contained in a stainless steel canister. The canister opening requires a positive seal weld, the properties and thickness of which are at least equal to those of the canister material. This paper describes the inertia friction welding process and a proposed equipment design concept that will provide a positive, reliable, inspectable, and full thickness seal weld while providing easily maintainable equipment, even though the weld is made in a highly contaminated hot cell. All studies and tests performed have shown the concept to be highly feasible. 2 refs., 6 figs

  14. Studies of waste-canister compatibility

    International Nuclear Information System (INIS)

    McCoy, H.E.

    1983-01-01

    Compatibility studies were conducted between 7 waste forms and 15 potential canister structural materials. The waste forms were Al-Si and Pb-Sn matrix alloys, FUETAP, glass, Synroc D, and waste particles coated with carbon or carbon plus silicon carbide. The canister materials included carbon steel (bare and with chromium or nickel coatings), copper, Monel, Cu-35% Ni, titanium (grades 2 and 12), several Inconels, aluminum alloy 5052, and two stainless steels. Tests of either 6888 or 8821 h were conducted at 100 and 300 0 C, which bracket the low and high limits expected during storage. Glass and FUETAP evolved sulfur, which reacted preferentially with copper, nickel, and alloys of these metals. The Pb-Sn matrix alloy stuck to all samples and the carbon-coated particles to most samples at 300 0 C, but the extent of chemical reaction was not determined. Testing for 0.5 h at 800 0 C was included because it is representative of a transportation accident and is required of casks containing nuclear materials. During these tests (1) glass and FUETAP evolved sulfur, (2) FUETAP evolved large amounts of gas, (3) Synroc stuck to titanium alloys, (4) glass was molten, and (5) both matrix alloys were molten with considerable chemical interactions with many of the canister samples. If this test condition were imposed on waste canisters, it would be design limiting in many waste storage concepts

  15. Inorganic analyses of volatilized and condensed species within prototypic Defense Waste Processing Facility (DWPF) canistered waste

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1992-01-01

    The high-level radioactive waste currently stored in carbon steel tanks at the Savannah River Site (SRS) will be immobilized in a borosilicate glass in the Defense Waste Processing Facility (DWPF). The canistered waste will be sent to a geologic repository for final disposal. The Waste Acceptance Preliminary Specifications (WAPS) require the identification of any inorganic phases that may be present in the canister that may lead to internal corrosion of the canister or that could potentially adversely affect normal canister handling. During vitrification, volatilization of mixed (Na, K, Cs)Cl, (Na, K, Cs) 2 SO 4 , (Na, K, Cs)BF 4 , (Na, K) 2 B 4 O 7 and (Na,K)CrO 4 species from glass melt condensed in the melter off-gas and in the cyclone separator in the canister pour spout vacuum line. A full-scale DWPF prototypic canister filled during Campaign 10 of the SRS Scale Glass Melter was sectioned and examined. Mixed (NaK)CI, (NaK) 2 SO 4 , (NaK) borates, and a (Na,K) fluoride phase (either NaF or Na 2 BF 4 ) were identified on the interior canister walls, neck, and shoulder above the melt pour surface. Similar deposits were found on the glass melt surface and on glass fracture surfaces. Chromates were not found. Spinel crystals were found associated with the glass pour surface. Reference amounts of the halides and sulfates were found retained in the glass and the glass chemistry, including the distribution of the halides and sulfates, was homogeneous. In all cases where rust was observed, heavy metals (Zn, Ti, Sn) from the cutting blade/fluid were present indicating that the rust was a reaction product of the cutting fluid with glass and heat sensitized canister or with carbon-steel contamination on canister interior. Only minimal water vapor is present so that internal corrosion of the canister, will not occur

  16. Thermal Predictions of the Cooling of Waste Glass Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2014-11-01

    Radioactive liquid waste from five decades of weapons production is slated for vitrification at the Hanford site. The waste will be mixed with glass forming additives and heated to a high temperature, then poured into canisters within a pour cave where the glass will cool and solidify into a stable waste form for disposal. Computer simulations were performed to predict the heat rejected from the canisters and the temperatures within the glass during cooling. Four different waste glass compositions with different thermophysical properties were evaluated. Canister centerline temperatures and the total amount of heat transfer from the canisters to the surrounding air are reported.

  17. Fracturing of simulated high-level waste glass in canisters

    International Nuclear Information System (INIS)

    Peters, R.D.; Slate, S.C.

    1981-09-01

    Waste-glass castings generated from engineering-scale developmental processes at the Pacific Northwest Laboratory are generally found to have significant levels of cracks. The causes and extent of fracturing in full-scale canisters of waste glass as a result of cooling and accidental impact are discussed. Although the effects of cracking on waste-form performance in a repository are not well understood, cracks in waste forms can potentially increase leaching surface area. If cracks are minimized or absent in the waste-glass canisters, the potential for radionuclide release from the canister package can be reduced. Additional work on the effects of cracks on leaching of glass is needed. In addition to investigating the extent of fracturing of glass in waste-glass canisters, methods to reduce cracking by controlling cooling conditions were explored. Overall, the study shows that the extent of glass cracking in full-scale, passively-cooled, continuous melting-produced canisters is strongly dependent on the cooling rate. This observation agrees with results of previously reported Pacific Northwest Laboratory experiments on bench-scale annealed canisters. Thus, the cause of cracking is principally bulk thermal stresses. Fracture damage resulting from shearing at the glass/metal interface also contributes to cracking, more so in stainless steel canisters than in carbon steel canisters. This effect can be reduced or eliminated with a graphite coating applied to the inside of the canister. Thermal fracturing can be controlled by using a fixed amount of insulation for filling and cooling of canisters. In order to maintain production rates, a small amount of additional facility space is needed to accomodate slow-cooling canisters. Alternatively, faster cooling can be achieved using the multi-staged approach. Additional development is needed before this approach can be used on full-scale (60-cm) canisters

  18. MCC-15: waste/canister accident testing and analysis method

    International Nuclear Information System (INIS)

    Slate, S.C.; Pulsipher, B.A.; Scott, P.A.

    1985-02-01

    The Materials Characterization Center (MCC) at the Pacific Northwest Laboratory (PNL) is developing standard tests to characterize the performance of nuclear waste forms under normal and accident conditions. As part of this effort, the MCC is developing MCC-15, Waste/Canister Accident Testing and Analysis. MCC-15 is used to test canisters containing simulated waste forms to provide data on the effects of accidental impacts on the waste form particle size and on canister integrity. The data is used to support the design of transportation and handling equipment and to demonstrate compliance with repository waste acceptance specifications. This paper reviews the requirements that led to the development of MCC-15, describes the test method itself, and presents some early results from tests on canisters representative of those proposed for the Defense Waste Processing Facility (DWPF). 13 references, 6 figures

  19. Storage and disposal of radioactive waste as glass in canisters

    International Nuclear Information System (INIS)

    Mendel, J.E.

    1978-12-01

    A review of the use of waste glass for the immobilization of high-level radioactive waste glass is presented. Typical properties of the canisters used to contain the glass, and the waste glass, are described. Those properties are used to project the stability of canisterized waste glass through interim storage, transportation, and geologic disposal

  20. 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

  1. Characterization of materials for waste-canister compatibility studies

    International Nuclear Information System (INIS)

    McCoy, H.E.; Mack, J.E.

    1981-10-01

    Sample materials of 7 waste forms and 15 potential canister materials were procured for compatibility tests. These materials were characterized before being placed in test, and the results are the main topic of this report. A test capsule was designed for the tests in which disks of a single waste form were contacted with duplicate samples of canister materials. The capsules are undergoing short-term tests at 800 0 C and long-term tests at 100 and 300 0 C

  2. 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

  3. Value Engineering Study for Closing Waste Packages Containing TAD Canisters

    International Nuclear Information System (INIS)

    Colleen Shelton-Davis

    2005-01-01

    The Office of Civilian Radioactive Waste Management announced their intention to have the commercial utilities package spent nuclear fuel in shielded, transportable, ageable, and disposable containers prior to shipment to the Yucca Mountain repository. This will change the conditions used as a basis for the design of the waste package closure system. The environment is now expected to be a low radiation, low contamination area. A value engineering study was completed to evaluate possible modifications to the existing closure system using the revised requirements. Four alternatives were identified and evaluated against a set of weighted criteria. The alternatives are (1) a radiation-hardened, remote automated system (the current baseline design); (2) a nonradiation-hardened, remote automated system (with personnel intervention if necessary); (3) a nonradiation-hardened, semi-automated system with personnel access for routine manual operations; and (4) a nonradiation-hardened, fully manual system with full-time personnel access. Based on the study, the recommended design is Alternative 2, a nonradiation-hardened, remote automated system. It is less expensive and less complex than the current baseline system, because nonradiation-hardened equipment can be used and some contamination control equipment is no longer needed. In addition, the inclusion of remote automation ensures throughput requirements are met, provides a more reliable process, and provides greater protection for employees from industrial accidents and radiation exposure than the semi-automated or manual systems. Other items addressed during the value engineering study as requested by OCRWM include a comparison to industry canister closure systems and corresponding lessons learned; consideration of closing a transportable, ageable, and disposable canister; and an estimate of the time required to perform a demonstration of the recommended closure system

  4. Value Engineering Study for Closing Waste Packages Containing TAD Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Colleen Shelton-Davis

    2005-11-01

    The Office of Civilian Radioactive Waste Management announced their intention to have the commercial utilities package spent nuclear fuel in shielded, transportable, ageable, and disposable containers prior to shipment to the Yucca Mountain repository. This will change the conditions used as a basis for the design of the waste package closure system. The environment is now expected to be a low radiation, low contamination area. A value engineering study was completed to evaluate possible modifications to the existing closure system using the revised requirements. Four alternatives were identified and evaluated against a set of weighted criteria. The alternatives are (1) a radiation-hardened, remote automated system (the current baseline design); (2) a nonradiation-hardened, remote automated system (with personnel intervention if necessary); (3) a nonradiation-hardened, semi-automated system with personnel access for routine manual operations; and (4) a nonradiation-hardened, fully manual system with full-time personnel access. Based on the study, the recommended design is Alternative 2, a nonradiation-hardened, remote automated system. It is less expensive and less complex than the current baseline system, because nonradiation-hardened equipment can be used and some contamination control equipment is no longer needed. In addition, the inclusion of remote automation ensures throughput requirements are met, provides a more reliable process, and provides greater protection for employees from industrial accidents and radiation exposure than the semi-automated or manual systems. Other items addressed during the value engineering study as requested by OCRWM include a comparison to industry canister closure systems and corresponding lessons learned; consideration of closing a transportable, ageable, and disposable canister; and an estimate of the time required to perform a demonstration of the recommended closure system.

  5. Decontamination of stainless steel canisters that contain high-level waste

    International Nuclear Information System (INIS)

    Bray, L.A.

    1987-01-01

    At the West Valley Nuclear Services Company (WVNSC) in West Valley, New York, high-level radioactive waste (HLW) will be vitrified into a borosilicate glass form and poured into large, stainless steel canisters. During the filling process, volatile fission products, principally 137 Cs, condense on the exterior of the canisters. The smearable contamination that remains on the canisters after they are filled and partially cooled must be removed from the canisters' exterior surfaces prior to their storage and ultimate shipment to a US Department of Energy (DOE) repository for disposal. A simple and effective method was developed for decontamination of HLW canisters. This method of chemical decontamination is applicable to a wide variety of contaminated equipment found in the nuclear industry. The process employs a reduction-oxidation system [Ce(III)/Ce(IV)] in nitric acid solution to chemically mill the surface of stainless steel, similar to the electropolishing process, but without the need for an applied electrical current. Contaminated canisters are simply immersed in the solution at controlled temperature and Ce(IV) concentration levels

  6. 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

  7. 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

  8. Transportation considerations related to waste forms and canisters for Defense TRU wastes

    International Nuclear Information System (INIS)

    Schneider, K.J.; Andrews, W.B.; Schreiber, A.M.; Rosenthal, L.J.; Odle, C.J.

    1981-09-01

    This report identifies and discusses the considerations imposed by transportation on waste forms and canisters for contact-handled, solid transuranic wastes from the US Department of Energy (DOE) activities. The report reviews (1) the existing raw waste forms and potential immobilized waste forms, (2) the existing and potential future DOE waste canisters and shipping containers, (3) regulations and regulatory trends for transporting commercial transuranic wastes on the ISA, (4) truck and rail carrier requirements and preferences for transporting the wastes, and (5) current and proposed Type B external packagings for transporting wastes

  9. Hanford Waste Vitrification Plant: Preliminary description of waste form and canister

    International Nuclear Information System (INIS)

    Mitchell, D.E.

    1986-01-01

    In July 1985, the US Department of Energy's Office of Civilian Radioactive Waste Management established the Waste Acceptance Process as the means by which defense high-level waste producers, such as the Hanford Waste Vitrification Plant, will develop waste acceptance requirements with the candidate geologic repositories. A complete description of the Waste Acceptance Process is contained in the Preliminary Hanford Waste Vitrification Plant Waste Form Qualification Plan. The Waste Acceptance Process defines three documents that high-level waste producers must prepare as a part of the process of assuming that a high-level waste product will be acceptable for disposal in a geologic repository. These documents are the Description of Waste Form and Canister, Waste Compliance Plan, and Waste Qualification Report. This document is the Hanford Waste Vitrification Plant Preliminary Description of Waste Form and Canister for disposal of Neutralized Current Acid Waste. The Waste Acceptance Specifications for the Hanford Waste Vitrification Plant have not yet been developed, therefore, this document has been structured to corresponds to the Waste Acceptance Preliminary Specifications for the Defense Waste Processing Facility High-Level Waste Form. Not all of the information required by these specifications is appropriate for inclusion in this Preliminary Description of Waste Form and Canister. Rather, this description is limited to information that describes the physical and chemical characteristics of the expected high-level waste form. The content of the document covers three major areas: waste form characteristics, canister characteristics, and canistered waste form characteristics. This information will be used by the candidate geologic repository projects as the basis for preliminary repository design activities and waste form testing. Periodic revisions are expected as the Waste Acceptance Process progresses

  10. Sediment mechanical response due to emplacement of a waste canister

    International Nuclear Information System (INIS)

    Karnes, C.H.; Dawson, P.R.; Silva, A.J.; Brown, W.T.

    1980-01-01

    Preliminary studies have been conducted to determine the interaction between a waste canister and seabed sediment during and after emplacement. Empirical and approximate methods for determining the depth reached by a freefall penetrator indicate that a boosted penetrator emplacement method may be necessary. Hole closure is necessary, but has not been verified because calculations and laboratory experiments show sensitivity to boundary conditions which control the degree of dynamic hole closure. Laboratory studies show that closure will take place by creep deformation but closure times in seabed environments are uncertain. For assumed thermomechanical properties of sediments, it is shown that a heat generating waste canister will probably not move a significant distancce during the heat generation period

  11. 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

  12. Acoustic monitoring techniques for corrosion degradation in cemented waste canisters

    International Nuclear Information System (INIS)

    Naish, C.C.; Buttle, D.; Wallace-Sims, R.; O'Brien, T.M.

    1991-01-01

    This report describes work carried out to investigate acoustic emission as a monitor of corrosion and degradation of wasteforms where the waste is potentially reactive metal. Electronic monitoring equipment has been designed, built and tested to allow long-term monitoring of a number of waste packages simultaneously. Acoustic monitoring experiments were made on a range of 1 litre cemented Magnox and aluminium samples cast into canisters comparing the acoustic events with hydrogen gas evolution rates and electrochemical corrosion rates. The attenuation of the acoustic signals by the cement grout under a range of conditions has been studied to determine the volume of wasteform that can be satisfactorily monitored by one transducer. The final phase of the programme monitored the acoustic events from full size (200 litre) cemented, inactive, simulated aluminium swarf wastepackages prepared at the AEA waste cementation plant at Winfrith. (Author)

  13. Friction Stir Welding of Copper Canisters for Nuclear Waste

    Energy Technology Data Exchange (ETDEWEB)

    Kaellgren, Therese

    2005-07-01

    The Swedish model for final disposal of nuclear fuel waste is based on copper canisters as a corrosion barrier with an inner pressure holding insert of cast iron. One of the methods to seal the copper canister is to use the Friction Stir Welding (FSW), a method invented by The Welding Institute (TWI). This work has been focused on characterisation of the FSW joints, and modelling of the process, both analytically and numerically. The first simulations were based on Rosenthal's analytical medium plate model. The model is simple to use, but has limitations. Finite element models were developed, initially with a two-dimensional geometry. Due to the requirements of describing both the heat flow and the tool movement, three-dimensional models were developed. These models take into account heat transfer, material flow, and continuum mechanics. The geometries of the models are based on the simulation experiments carried out at TWI and at Swedish Nuclear Fuel Waste and Management Co (SKB). Temperature distribution, material flow and their effects on the thermal expansion were predicted for a full-scale canister and lid. The steady state solutions have been compared with temperature measurements, showing good agreement. Microstructure and hardness profiles have been investigated by optical microscope, Scanning Electron Microscope (SEM), Electron Back Scatter Diffraction (EBSD) and Rockwell hardness measurements. EBSD visualisation has been used to determine the grain size distribution and the appearance of twins and misorientation within grains. The orientation maps show a fine uniform equiaxed grain structure. The root of the weld exhibits the smallest grains and many annealing twins. This may be due to deformation after recrystallisation. The appearance of the nugget and the grain size depends on the position of the weld. A large difference can be seen both in hardness and grain size between the start of the weld and when the steady state is reached.

  14. Annular air space effects on nuclear waste canister temperatures in a deep geologic waste repository

    International Nuclear Information System (INIS)

    Lowry, W.E.; Cheung, H.; Davis, B.W.

    1980-01-01

    Air spaces in a deep geologic repository for nuclear high level waste will have an important effect on the long-term performance of the waste package. The important temperature effects of an annular air gap surrounding a high level waste canister are determined through 3-D numerical modeling. Air gap properties and parameters specifically analyzed and presented are the air gap size, surfaces emissivity, presence of a sleeve, and initial thermal power generation rate; particular emphasis was placed on determining the effect of these variables have on the canister surface temperature. Finally a discussion based on modeling results is presented which specifically relates the results to NRC regulatory considerations

  15. Corrosion of canister materials for radioactive waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Kienzler, Bernhard [KIT Karlsruhe (Germany). Institut fuer Nukleare Entsorgung (INE)

    2017-08-15

    In the period between 1980 and 2004, corrosion studies on various metallic materials have been performed at the Research Center Karlsruhe. The objectives of these experimental studies addressed mainly the performance of canister materials for heat producing, high-level wastes and spent nuclear fuels for a repository in a German salt dome. Additional studies covered the performance of steels for packaging wastes with negligible heat production under conditions to be expected in rocksalt and in the Konrad iron ore mine. The results of the investigations have been published in journals and conference proceedings but also in ''grey literature''. This paper presents a summary of the results of corrosion experiments with fine-grained steels and nodular cast steel.

  16. Corrosion of canister materials for radioactive waste disposal

    International Nuclear Information System (INIS)

    Kienzler, Bernhard

    2017-01-01

    In the period between 1980 and 2004, corrosion studies on various metallic materials have been performed at the Research Center Karlsruhe. The objectives of these experimental studies addressed mainly the performance of canister materials for heat producing, high-level wastes and spent nuclear fuels for a repository in a German salt dome. Additional studies covered the performance of steels for packaging wastes with negligible heat production under conditions to be expected in rocksalt and in the Konrad iron ore mine. The results of the investigations have been published in journals and conference proceedings but also in ''grey literature''. This paper presents a summary of the results of corrosion experiments with fine-grained steels and nodular cast steel.

  17. The remote handling of canisters containing nuclear waste in glass at the Savannah River Plant

    International Nuclear Information System (INIS)

    Callan, J.E.

    1986-01-01

    The Defense Waste Processing Facility (DWPF) is a complete production area being constructed at the Savannah River Plant for the immobilization of nuclear waste in glass. The remote handling of canisters filled with nuclear waste in glass is an essential part of the process of the DWPF at the Savannah River Plant. The canisters are filled with nuclear waste containing up to 235,000 curies of radioactivity. Handling and movement of these canisters must be accomplished remotely since they radiate up to 5000 R/h. Within the Vitrification Building during filling, cleaning, and sealing, canisters are moved using standard cranes and trolleys and a specially designed grapple. During transportation to the Glass Waste Storage Building, a one-of-a-kind, specially designed Shielded Canister Transporter (SCT) is used. 8 figs

  18. Copper canisters for nuclear high level waste disposal. Corrosion aspects

    International Nuclear Information System (INIS)

    Werme, L.; Sellin, P.; Kjellbert, N.

    1992-10-01

    A corrosion analysis of a thick-walled copper canister for spent fuel disposal is discussed. The analysis has shown that there are no rapid mechanisms that may lead to canister failure, indicating an anticipated corrosion service life of several millions years. If further analysis of the copper canister is considered, it should be concentrated on identifying and evaluating processes other than corrosion, which may have a potential for leading to canister failure. (au)

  19. Molecular Contamination on Anodized Aluminum Components of the Genesis Science Canister

    Science.gov (United States)

    Burnett, D. S.; McNamara, K. M.; Jurewicz, A.; Woolum, D.

    2005-01-01

    Inspection of the interior of the Genesis science canister after recovery in Utah, and subsequently at JSC, revealed a darkening on the aluminum canister shield and other canister components. There has been no such observation of film contamination on the collector surfaces, and preliminary spectroscopic ellipsometry measurements support the theory that the films observed on the anodized aluminum components do not appear on the collectors to any significant extent. The Genesis Science Team has made an effort to characterize the thickness and composition of the brown stain and to determine if it is associated with molecular outgassing.Detailed examination of the surfaces within the Genesis science canister reveals that the brown contamination is observed to varying degrees, but only on surfaces exposed in space to the Sun and solar wind hydrogen. In addition, the materials affected are primarily composed of anodized aluminum. A sharp line separating the sun and shaded portion of the thermal closeout panel is shown. This piece was removed from a location near the gold foil collector within the canister. Future plans include a reassembly of the canister components to look for large-scale patterns of contamination within the canister to aid in revealing the root cause.

  20. Criticality safety calculations for the nuclear waste disposal canisters

    International Nuclear Information System (INIS)

    Anttila, M.

    1996-12-01

    The criticality safety of the copper/iron canisters developed for the final disposal of the Finnish spent fuel has been studied with the MCNP4A code based on the Monte Carlo technique and with the fuel assembly burnup programs CASMO-HEX and CASMO-4. Two rather similar types of spent fuel disposal canisters have been studied. One canister type has been designed for hexagonal VVER-440 fuel assemblies used at the Loviisa nuclear power plant (IVO canister) and the other one for square BWR fuel bundles used at the Olkiluoto nuclear power plant (TVO canister). (10 refs.)

  1. 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

  2. The design analysis of ACP-canister for nuclear waste disposal

    International Nuclear Information System (INIS)

    Raiko, H.

    1992-05-01

    The design basis, dimensioning and some manufacturing aspects of the Advanced Cold Process Canister (ACPC) for the nuclear waste disposal is summarized in the report. The strength of the canister has been evaluated in normal design load condition and in extreme high hydrostatic pressure load condition possibly caused by ice age (orig.)

  3. Defense Waste Processing Facility Canister Closure Weld Current Validation Testing

    Energy Technology Data Exchange (ETDEWEB)

    Korinko, P. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Maxwell, D. N. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2018-01-29

    Two closure welds on filled Defense Waste Processing Facility (DWPF) canisters failed to be within the acceptance criteria in the DWPF operating procedure SW4-15.80-2.3 (1). In one case, the weld heat setting was inadvertently provided to the canister at the value used for test welds (i.e., 72%) and this oversight produced a weld at a current of nominally 210 kA compared to the operating procedure range (i.e., 82%) of 240 kA to 263 kA. The second weld appeared to experience an instrumentation and data acquisition upset. The current for this weld was reported as 191 kA. Review of the data from the Data Acquisition System (DAS) indicated that three of the four current legs were reading the expected values, approximately 62 kA each, and the fourth leg read zero current. Since there is no feasible way by further examination of the process data to ascertain if this weld was actually welded at either the target current or the lower current, a test plan was executed to provide assurance that these Nonconforming Welds (NCWs) meet the requirements for strength and leak tightness. Acceptance of the welds is based on evaluation of Test Nozzle Welds (TNW) made specifically for comparison. The TNW were nondestructively and destructively evaluated for plug height, heat tint, ultrasonic testing (UT) for bond length and ultrasonic volumetric examination for weld defects, burst pressure, fractography, and metallography. The testing was conducted in agreement with a Task Technical and Quality Assurance Plan (TTQAP) (2) and applicable procedures.

  4. NAC gets OK for waste canister, looks for buyers

    International Nuclear Information System (INIS)

    Newman, P.

    1994-01-01

    The Nuclear Regulatory Commission has given design approval to the first dual-purpose waste cansiter suitable for storing and transporting irradiated nuclear fuel. The cask could be commercially available by January 1995. NRC issued a transportation certificate for the canister, which was developed by Atlanta-based NAC Services Inc., a subsidiary of NAC Holding Inc. That certificate, which says the cask is a suitable vessel for transporting radioactive wastes by rail and truck, is the first credential of a two-part licensing process the design must acquire. Testing of the cask has been extensive, including drop tests and pin-puncture tests. Roughly 19 feet long and eight feet in diameter, the cask is designed to hold 26 pressurized water reactor fuel assemblies. NAC officials say the cask design will soon be adapted to accomodate larger boiling water reactor fuel assemblies. Utilities will need some convincing that the dual-purpose $1.5 million cask is worth the money, particularly since companies currently have no use for the cask's transportation capabilities

  5. Remote handling of canisters containing nuclear waste in glass at the Savannah River Plant

    International Nuclear Information System (INIS)

    Callan, J.E.

    1986-01-01

    The Defense Waste Processing Facility is being constructed at the Savannah River Plant at a cost of $870 million to immobilize the defense high-level radioactive waste. This radioactive waste is being added to borosilicate glass for later disposal in a federal repository. The borosilicate glass is poured into stainless steel canisters for storage. These canisters must be handled remotely because of their high radioactivity, up to 5000 R/h. After the glass has been poured into the canister which will be temporarily sealed, it is transferred to a decontamination cell and decontaminated. The canister is then transferred to the weld cell where a permanent cap is welded into place. The canisters must then be transported from the processing building to a storage vault on the plant until the federal repository is available. A shielded canister transporter (SCT) has been designed and constructed for this purpose. The design of the SCT vehicle allows the safe transport of a highly radioactive canister containing borosilicate glass weighing 2300 kg with a radiation level up to 5000 R/h from one building to another. The design provides shielding for the operator in the cab of the vehicle to be below 0.5 rem/h. The SCT may also be used to load the final shipping cask when the federal repository is ready to receive the canisters

  6. Commercial radioactive waste management system feasibility with the universal canister concept. Volume 1

    International Nuclear Information System (INIS)

    Morissette, R.P.; Schneringer, P.E.; Lane, R.K.; Moore, R.L.; Young, K.A.

    1986-01-01

    A Program Research and Development Announcement (PRDA) was initiated by DOE to solicit from industry new and novel ideas for improvements in the nuclear waste management system. GA Technologies Inc. was contracted to study a system utilizing a universal canister which could be loaded at the reactor and used throughout the waste management system. The proposed canister was developed with the objective of meeting the mission requirements with maximum flexibility and at minimum cost. Canister criteria were selected from a thorough analysis of the spent fuel inventory, and canister concepts were evaluated along with the shipping and storage casks to determine the maximum payload. Engineering analyses were performed on various cask/canister combinations. One important criterion was the interchangeability of the canisters between truck and rail cask systems. A canister was selected which could hold three PWR intact fuel elements or up to eight consolidated PWR fuel elements. One canister could be shipped in an overweight truck cask or six in a rail cask. Economic analysis showed a cost savings of the reference system under consideration at that time

  7. 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

  8. 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.

  9. 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

  10. Examining the role of canister cooling conditions on the formation of nepheline from nuclear waste glasses

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

    Nepheline (NaAlSiO₄) crystals can form during slow cooling of high-level waste (HLW) glass after it has been poured into a waste canister. Formation of these crystals can adversely affect the chemical durability of the glass. The tendency for nepheline crystallization to form in a HLW glass increases with increasing concentrations of Al₂O₃ and Na₂O.

  11. Effects of stabilizers on the heat transfer characteristics of a nuclear waste canister

    International Nuclear Information System (INIS)

    Vafai, K.; Ettefagh, J.

    1986-07-01

    This report summarizes the feasibility and the effectiveness of using stabilizers (internal metal structural components) to augment the heat transfer characteristics of a nuclear waste canister. The problem was modeled as a transient two-dimensional heat transfer in two physical domains - the stabilizer and the wedge (a 30-degree-angle canister segment), which includes the heat-producing spent-fuel rods. This problem is solved by a simultaneous and interrelated numerical investigation of the two domains in cartesian and polar coordinate systems. The numerical investigations were performed for three cases. In the first case, conduction was assumed to be the dominant mechanism for heat transfer. The second case assumed that radiation was the dominant mechanism, and in the third case both radiation and conduction were considered as mechanisms of heat transfer. The results show that for typical conditions in a waste package design, the stabilizers are quite effective in reducing the overall temperature in a waste canister. Furthermore, the results show that increasing the stabilizer thickness over the thickness specified in the present design has a negligible effect on the temperature distribution in the canister. Finally, the presence of the stabilizers was found to shift the location of the peak temperature areas in the waste canister

  12. Canister materials proposed for final disposal of high level nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Mattson, E; Odoj, R; Merz, E [eds.

    1981-06-01

    The nuclear waste will be enclosed in corrosion resistant canisters. These will be deposited in repositories in geological formations, such as granite, basalt, clay, bedded or domed salt, or the sediments beneath the deep ocean floor. There the canisters will be exposed to groundwater, brine or seawater at an elevated temperature. Species formed by radiolysis may effect the corrosivity of the agent. The corrosion resistance of candidate canister materials is evaluated by corrosion tests and by thermodynamic and mass transport calculations. Examinations of ancient metal objects after long exposure in nature may give additional information. On the basis of the work carried out so far, the principal candidate canister materials are titanium materials, copper, and highpurity alumina.

  13. Iron-nickel alloys as canister material for radioactive waste disposal in underground repositories

    International Nuclear Information System (INIS)

    Apps, J.A.

    1982-01-01

    Canisters containing high-level radioactive waste must retain their integrity in an underground waste repository for at least one thousand years after burial (Nuclear Regulatory Commission, 1981). Since no direct means of verifying canister integrity is plausible over such a long period, indirect methods must be chosen. A persuasive approach is to examine the natural environment and find a suitable material which is thermodynamically compatible with the host rock under the environmental conditions with the host rock under the environmental conditions expected in a waste repository. Several candidates have been proposed, among them being iron-nickel alloys that are known to occur naturally in altered ultramafic rocks. The following review of stability relations among iron-nickel alloys below 350 0 C is the initial phase of a more detailed evaluation of these alloys as suitable canister materials

  14. Dew point, internal gas pressure, and chemical composition of the gas within the free volume of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    Harbour, J.R.; Herman, D.T.; Crump, S.; Miller, T.J.; McIntosh, J.

    1996-01-01

    The Defense Waste Processing Facility (DWPF) produced 55 canistered waste forms containing simulated waste glass during the four Waste Qualification campaigns of the DWPF Startup Test Program. Testing of the gas within the free volume of these canisters for dew point, internal gas pressure, and chemical composition was performed as part of a continuing effort to demonstrate compliance with the Waste Acceptance Product Specifications. Results are presented for six glass-filled canisters. The dew points within the canisters met the acceptance criterion of < 20 degrees C for all six canisters. Factors influencing the magnitude of the dew point are presented. The chemical composition of the free volume gas was indistinguishable from air for all six canisters. Hence, no foreign materials were present in the gas phase of these canisters. The internal gas pressures within the sealed canisters were < 1 atm at 25 degrees C for all six canisters which readily met the acceptance criterion of an internal gas pressure of less than 1.5 atm at 25 degrees C. These results provided the evidence required to demonstrate compliance with the Waste Acceptance Product Specifications

  15. The Characteristics of Welding Joint on Stainless Steel as a Candidate of High Level Waste Canister

    International Nuclear Information System (INIS)

    Aisyah; Herlan-Martono

    2000-01-01

    High level waste is the waste generated from reprocessing of the spent fuels. This type of waste is vitrified with borosilicate glass to become waste-glass. This waste glass is contained in a canister made of austenitic stainless steel. The canister material is subjected to be welded during fabrication and utilization. The character of the welding joint that is the function of the electrical current used in the welding process have been studied. The strength of the joint is tested mechanically i.e.: the tensile strength and hardness test. The result shows that the higher the current used in welding process, the better the strength of the joint and as well the tensile strength. The optimum current is 110 A. From the hardness test, it was figured that the length of the HAZ area is 14 mm. The material in HAZ area is the hardest compared to the others, it is due to the appearance of the chrome-carbide. The welding of the canister with such a condition, during fabrication as well as during the utilization of the canister for the container of the high level waste with the PWHT process gives better result. (author)

  16. A review of the possible effects of hydrogen on lifetime of carbon steel nuclear waste canisters

    International Nuclear Information System (INIS)

    Turnbull, A.

    2009-07-01

    In Switzerland, the National Cooperative for the Disposal of Radioactive Waste (Nagra) is responsible for developing an effective method for the safe disposal of vitrified high level waste (HLW) and spent fuel. One of the options for disposal canisters is thick-walled carbon steel. The canisters, which would have a diameter of about 1 m and a length of about 3 m (HLW) or about 5 m (spent fuel), will be embedded in horizontal tunnels and surrounded with bentonite clay. The regulatory requirement for the minimum canister lifetime is 1000 years but demonstration of a minimum lifetime of 10,000 years would be desirable. The pore-water to which the canister will be exposed is of marine origin with about 0.1-0.3 M Cl-. Since hydrogen is generated during the corrosion process, it is necessary to assess the probability of hydrogen assisted cracking modes and to make recommendations to eliminate that probability. To that aim, key reports detailing projections for the local environment and associated corrosion rate of the waste canister have been evaluated with the focus on the implication for the absorbed hydrogen concentration in the steel. Simple calculations of hydrogen diffusion and accumulation in the inner compartment of the sealed canister indicate that a pressure equivalent to that for gas pockets external to the canister (envisaged to be about 10 MPa) may be attained in the proposed exposure time, an important consideration since it is not possible to modify the internal surface of the closure weld. Current ideas on mechanisms of hydrogen assisted cracking are assessed from which it is concluded that the mechanistic understanding and associated models of hydrogen assisted cracking are insufficient to provide a framework for quantitative prediction for this application. The emphasis then was to identify threshold conditions for cracking and to evaluate the likelihood that these may be exceeded over the lifetime of the containment. Based on an analysis of data in the

  17. Canister design concepts for disposal of spent fuel and high level waste

    Energy Technology Data Exchange (ETDEWEB)

    Patel, R.; Punshon, C.; Nicholas, J.; Bastid, P.; Zhou, R.; Schneider, C.; Bagshaw, N.; Howse, D.; Hutchinson, E. [TWI Ltd, Cambridge, (United Kingdom); Asano, R. [Hitachi Zosen Corporation, Osaka (Japan); King, S. [Integrity Corrosion Consulting Ltd, Calgary, Alberta (Canada)

    2012-10-15

    As part of its long-term plans for development of a repository for spent fuel (SF) and high level waste (HLW), Nagra is exploring various options for the selection of materials and design concepts for disposal canisters. The selection of suitable canister options is driven by a series of requirements, one of the most important of which is providing a minimum 1000 year lifetime without breach of containment. One candidate material is carbon steel, because of its relatively low corrosion rate under repository conditions and because of the advanced state of overall technical maturity related to construction and fabrication. Other materials and design options are being pursued in parallel studies. The objective of the present study was to develop conceptual designs for carbon steel SF and HLW canisters along with supporting justification. The design process and outcomes result in design concepts that deal with all key aspects of canister fabrication, welding and inspection, short-term performance (handling and emplacement) and long-term performance (corrosion and structural behaviour after disposal). A further objective of the study is to use the design process to identify the future work that is required to develop detailed designs. The development of canister designs began with the elaboration of a number of design requirements that are derived from the need to satisfy the long-term safety requirements and the operational safety requirements (robustness needed for safe handling during emplacement and potential retrieval). It has been assumed based on radiation shielding calculations that the radiation dose rate at the canister surfaces will be at a level that prohibits manual handling, and therefore a hot cell and remote handling will be needed for filling the canisters and for final welding operations. The most important canister requirements were structured hierarchically and set in the context of an overall design methodology. Conceptual designs for SF canisters

  18. Canister design concepts for disposal of spent fuel and high level waste

    International Nuclear Information System (INIS)

    Patel, R.; Punshon, C.; Nicholas, J.; Bastid, P.; Zhou, R.; Schneider, C.; Bagshaw, N.; Howse, D.; Hutchinson, E.; Asano, R.; King, S.

    2012-10-01

    As part of its long-term plans for development of a repository for spent fuel (SF) and high level waste (HLW), Nagra is exploring various options for the selection of materials and design concepts for disposal canisters. The selection of suitable canister options is driven by a series of requirements, one of the most important of which is providing a minimum 1000 year lifetime without breach of containment. One candidate material is carbon steel, because of its relatively low corrosion rate under repository conditions and because of the advanced state of overall technical maturity related to construction and fabrication. Other materials and design options are being pursued in parallel studies. The objective of the present study was to develop conceptual designs for carbon steel SF and HLW canisters along with supporting justification. The design process and outcomes result in design concepts that deal with all key aspects of canister fabrication, welding and inspection, short-term performance (handling and emplacement) and long-term performance (corrosion and structural behaviour after disposal). A further objective of the study is to use the design process to identify the future work that is required to develop detailed designs. The development of canister designs began with the elaboration of a number of design requirements that are derived from the need to satisfy the long-term safety requirements and the operational safety requirements (robustness needed for safe handling during emplacement and potential retrieval). It has been assumed based on radiation shielding calculations that the radiation dose rate at the canister surfaces will be at a level that prohibits manual handling, and therefore a hot cell and remote handling will be needed for filling the canisters and for final welding operations. The most important canister requirements were structured hierarchically and set in the context of an overall design methodology. Conceptual designs for SF canisters

  19. Canister Design for Deep Borehole Disposal of Nuclear Waste (CD-ROM)

    National Research Council Canada - National Science Library

    Hoag, Christopher I

    2006-01-01

    ...: 1 CD-ROM; 4 3/4 in.; 28.7 MB. ABSTRACT: The objective of this thesis was to design a canister for the disposal of spent nuclear fuel and other high-level waste in deep borehole repositories using currently available and proven oil, gas...

  20. 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

  1. Yucca Mountain project canister material corrosion studies as applied to the electrometallurgical treatment metallic waste form

    International Nuclear Information System (INIS)

    Keiser, D.D.

    1996-11-01

    Yucca Mountain, Nevada is currently being evaluated as a potential site for a geologic repository. As part of the repository assessment activities, candidate materials are being tested for possible use as construction materials for waste package containers. A large portion of this testing effort is focused on determining the long range corrosion properties, in a Yucca Mountain environment, for those materials being considered. Along similar lines, Argonne National Laboratory is testing a metallic alloy waste form that also is scheduled for disposal in a geologic repository, like Yucca Mountain. Due to the fact that Argonne's waste form will require performance testing for an environment similar to what Yucca Mountain canister materials will require, this report was constructed to focus on the types of tests that have been conducted on candidate Yucca Mountain canister materials along with some of the results from these tests. Additionally, this report will discuss testing of Argonne's metal waste form in light of the Yucca Mountain activities

  2. TOUGH - a numerical model for nonisothermal unsaturated flow to study waste canister heating effects

    International Nuclear Information System (INIS)

    Pruess, K.; Wang, J.S.Y.

    1984-01-01

    The physical processes modeled and the mathematical and numerical methods employed in a simulator for non-isothermal flow of water, vapor, and air in permeable media are briefly summarized. The simulator has been applied to study thermohydrological conditions in the near vicinity of high-level nuclear waste packages emplaced in unsaturated rocks. The studies reported here specifically address the question whether or not the waste canister environment will dry up in the thermal phase. 13 references, 8 figures, 2 tables

  3. Enhanced Earthquake-Resistance on the High Level Radioactive Waste Canister

    International Nuclear Information System (INIS)

    Choi, Youngchul; Yoon, Chanhoon; Lee, Jeaowan; Kim, Jinsup; Choi, Heuijoo

    2014-01-01

    In this paper, the earthquake-resistance type buffer was developed with the method protecting safely about the earthquake. The main parameter having an effect on the earthquake-resistant performance was analyzed and the earthquake-proof type buffer material was designed. The shear analysis model was developed and the performance of the earthquake-resistance buffer material was evaluated. The dynamic behavior of the radioactive waste disposal canister was analyzed in case the earthquake was generated. In the case, the disposal canister gets the serious damage. In this paper, the earthquake-resistance buffer material was developed in order to prevent this damage. By putting the buffer in which the density is small between the canister and buffer, the earthquake-resistant performance was improved about 80%

  4. Tritium Packages and 17th RH Canister Categories of Transuranic Waste Stored Below Ground within Area G

    Energy Technology Data Exchange (ETDEWEB)

    Hargis, Kenneth Marshall [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-01

    A large wildfire called the Las Conchas Fire burned large areas near Los Alamos National Laboratory (LANL) in 2011 and heightened public concern and news media attention over transuranic (TRU) waste stored at LANL’s Technical Area 54 (TA-54) Area G waste management facility. The removal of TRU waste from Area G had been placed at a lower priority in budget decisions for environmental cleanup at LANL because TRU waste removal is not included in the March 2005 Compliance Order on Consent (Reference 1) that is the primary regulatory driver for environmental cleanup at LANL. The Consent Order is a settlement agreement between LANL and the New Mexico Environment Department (NMED) that contains specific requirements and schedules for cleaning up historical contamination at the LANL site. After the Las Conchas Fire, discussions were held by the U.S. Department of Energy (DOE) with the NMED on accelerating TRU waste removal from LANL and disposing it at the Waste Isolation Pilot Plant (WIPP). This report summarizes available information on the origin, configuration, and composition of the waste containers within the Tritium Packages and 17th RH Canister categories; their physical and radiological characteristics; the results of the radioassays; and potential issues in retrieval and processing of the waste containers.

  5. Analysis of Welding Joint on Handling High Level Waste-Glass Canister

    International Nuclear Information System (INIS)

    Herlan Martono; Aisyah; Wati

    2007-01-01

    The analysis of welding joint of stainless steel austenitic AISI 304 for canister material has been studied. At the handling of waste-glass canister from melter below to interim storage, there is a step of welding of canister lid. Welding quality must be kept in a good condition, in order there is no gas out pass welding pores and canister be able to lift by crane. Two part of stainless steel plate in dimension (200 x 125 x 3) mm was jointed by welding. Welding was conducted by TIG machine with protection gas is argon. Electric current were conducted for welding were 70, 80, 90, 100, 110, 120, 130, and 140 A. Welded plates were cut with dimension according to JIS 3121 standard for tensile strength test. Hardness test in welding zone, HAZ, and plate were conducted by Vickers. Analysis of microstructure by optic microscope. The increasing of electric current at the welding, increasing of tensile strength of welding yields. The best quality welding yields using electric current was 110 A. At the welding with electric current more than 110 A, the electric current influence towards plate quality, so that decreasing of stainless steel plate quality and breaking at the plate. Tensile strength of stainless steel plate welding yields in requirement conditions according to application in canister transportation is 0.24 kg/mm 2 . (author)

  6. Recommendations for codes and standards to be used for design and fabrication of high level waste canister

    International Nuclear Information System (INIS)

    Bermingham, A.J.; Booker, R.J.; Booth, H.R.; Ruehle, W.G.; Shevekov, S.; Silvester, A.G.; Tagart, S.W.; Thomas, J.A.; West, R.G.

    1978-01-01

    This study identifies codes, standards, and regulatory requirements for developing design criteria for high-level waste (HLW) canisters for commercial operation. It has been determined that the canister should be designed as a pressure vessel without provision for any overpressure protection type devices. It is recommended that the HLW canister be designed and fabricated to the requirements of the ASME Section III Code, Division 1 rules, for Code Class 3 components. Identification of other applicable industry and regulatory guides and standards are provided in this report. Requirements for the Design Specification are found in the ASME Section III Code. It is recommended that design verification be conducted principally with prototype testing which will encompass normal and accident service conditions during all phases of the canister life. Adequacy of existing quality assurance and licensing standards for the canister was investigated. One of the recommendations derived from this study is a requirement that the canister be N stamped. In addition, acceptance standards for the HLW waste should be established and the waste qualified to those standards before the canister is sealed. A preliminary investigation of use of an overpack for the canister has been made, and it is concluded that the use of an overpack, as an integral part of overall canister design, is undesirable, both from a design and economics standpoint. However, use of shipping cask liners and overpack type containers at the Federal repository may make the canister and HLW management safer and more cost effective. There are several possible concepts for canister closure design. These concepts can be adapted to the canister with or without an overpack. A remote seal weld closure is considered to be one of the most suitable closure methods; however, mechanical seals should also be investigated

  7. 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)

  8. Cost Comparison for the Transfer of Select Calcined Waste Canisters to the Monitored Geologic Repository at Yucca Mountain, NV

    International Nuclear Information System (INIS)

    Michael B. Heiser; Clark B. Millet

    2005-01-01

    This report performs a life-cycle cost comparison of three proposed canister designs for the shipment and disposition of Idaho National Laboratory high-level calcined waste currently in storage at the Idaho Nuclear Technology and Engineering Center to the proposed national monitored geologic repository at Yucca Mountain, Nevada. Concept A (2 x 10-ft) and Concept B (2 x 15-ft) canisters are comparable in design, but they differ in size and waste loading options and vary proportionally in weight. The Concept C (5.5 x 17.5-ft) canister (also called the ''super canister''), while similar in design to the other canisters, is considerably larger and heavier than Concept A and B canisters and has a greater wall thickness. This report includes estimating the unique life-cycle costs for the three canister designs. Unique life-cycle costs include elements such as canister purchase and filling at the Idaho Nuclear Technology and Engineering Center, cask preparation and roundtrip consignment costs, final disposition in the monitored geologic repository (including canister off-loading and placement in the final waste disposal package for disposition), and cask purchase. Packaging of the calcine ''as-is'' would save $2.9 to $3.9 billion over direct vitrification disposal in the proposed national monitored geologic repository at Yucca Mountain, Nevada. Using the larger Concept C canisters would use 0.75 mi less of tunnel space, cost $1.3 billion less than 10-ft canisters of Concept A, and would be complete in 6.2 years

  9. Research of radioactive waste storage cask/canister materials, spent nuclear fuels and various radioactive waste forms and development of their assessment methods. Final report for Stage 3

    International Nuclear Information System (INIS)

    Dobrev, D.; Balek, V.; Červinka, R.; Večerník, P.; Člupek, M.; Kouřil, M.; Novák, P.; Stoulil, J.; Silber, R.

    2013-08-01

    The main topics treated are: Research and development of methodologies for canister/cask material degradation assessment; Laboratory research of selected materials of canister/cask with radioactive waste; and Research and assessment of canister/cask materials in natural granite rocks. Two additional documents are appended: Corrosion rate determination for samples in compacted bentonite in anaerobic conditions (methodology), and Roll test for corrosion test in an occluded solution at the interface between a radioactive waste disposal canister and the bentonite cover. (P.A.)

  10. Friction stir welding - an alternative method for sealing nuclear waste storage canisters

    Energy Technology Data Exchange (ETDEWEB)

    Andrews, R.E. [TWI Ltd, Cambridge (United Kingdom)

    2004-12-01

    When welding 50 mm thick copper a very high heat input is required to combat the high thermal diffusivity and only the Electron Beam Welding (EBW) process had this capability when this copper canister concept was conceived. Despite the encouraging results achieved using EBW with thick section copper, SKB felt that it would be prudent to assess other joining methods. This assessment concluded that friction welding, could also provide very high quality welds to satisfy the service life requirements of the SKB canister design. A friction welding variant called Friction Stir Welding (FSW) was shown to have the capability of welding 3 mm thick copper sheet with excellent integrity and reproducibility. This later provided sufficient encouragement for SKB to consider the potential of FSW as a method for joining thick section copper, using relatively simple machine tool based technology. It was thought that FSW might provide an alternative or complementary method for welding lids, or bases to canisters. In 1997 an FSW development programme started at TWI, focussed on the feasibility of welding 10 mm thick copper plate. Once this task was successfully completed, work continued to demonstrate that progressively thicker plate, up to 50 mm thick, could be joined. At this stage, with process viability established, a full size experimental FSW canister machine was designed and built. Work with this machine finished in January 2003, when it had been shown that FSW could definitely be used to weld lids to full size canisters. This report summarises the TWI development of FSW for SKB from 1997 to January 2003. It also highlights the important aspects of the process and the project milestones that will help to ensure that SKB has a welding technology that can be used with confidence for production fabrication of copper waste storage canisters in the future. The overall conclusion to this FSW development is that there is no doubt that the FSW process could be used to produce full

  11. Friction stir welding - an alternative method for sealing nuclear waste storage canisters

    International Nuclear Information System (INIS)

    Andrews, R.E.

    2004-12-01

    When welding 50 mm thick copper a very high heat input is required to combat the high thermal diffusivity and only the Electron Beam Welding (EBW) process had this capability when this copper canister concept was conceived. Despite the encouraging results achieved using EBW with thick section copper, SKB felt that it would be prudent to assess other joining methods. This assessment concluded that friction welding, could also provide very high quality welds to satisfy the service life requirements of the SKB canister design. A friction welding variant called Friction Stir Welding (FSW) was shown to have the capability of welding 3 mm thick copper sheet with excellent integrity and reproducibility. This later provided sufficient encouragement for SKB to consider the potential of FSW as a method for joining thick section copper, using relatively simple machine tool based technology. It was thought that FSW might provide an alternative or complementary method for welding lids, or bases to canisters. In 1997 an FSW development programme started at TWI, focussed on the feasibility of welding 10 mm thick copper plate. Once this task was successfully completed, work continued to demonstrate that progressively thicker plate, up to 50 mm thick, could be joined. At this stage, with process viability established, a full size experimental FSW canister machine was designed and built. Work with this machine finished in January 2003, when it had been shown that FSW could definitely be used to weld lids to full size canisters. This report summarises the TWI development of FSW for SKB from 1997 to January 2003. It also highlights the important aspects of the process and the project milestones that will help to ensure that SKB has a welding technology that can be used with confidence for production fabrication of copper waste storage canisters in the future. The overall conclusion to this FSW development is that there is no doubt that the FSW process could be used to produce full

  12. Automated waste canister docking and emplacement using a sensor-based intelligent controller

    International Nuclear Information System (INIS)

    Drotning, W.D.

    1992-08-01

    A sensor-based intelligent control system is described that utilizes a multiple degree-of-freedom robotic system for the automated remote manipulation and precision docking of large payloads such as waste canisters. Computer vision and ultrasonic proximity sensing are used to control the automated precision docking of a large object with a passive target cavity. Real-time sensor processing and model-based analysis are used to control payload position to a precision of ± 0.5 millimeter

  13. Cost analysis for application of solidified waste fission product canisters in U.S. Army steam plants

    International Nuclear Information System (INIS)

    Sande, W.E.; Bjorklund, W.J.; Brooks, N.A.

    1977-04-01

    The main objectives of the present study are to design steam plants using projected waste fission product canister characteristics, to analyze the overall impact and cost/benefit to the nuclear fuel cycle associated with these plants, and to develop plans for this application if the cost analysis so warrants it. The construction and operation of a steam plant fueled with waste fission product canisters would require the involvement and cooperation of various government agencies and private industry; thus the philosophies of these groups were studied. These philosophies are discussed, followed by a forecast of canister supply, canister characteristics, and strategies for Army canister use. Another section describes the safety and licensing of these steam plants since this affects design and capital costs. The discussion of steam plant design includes boiler concepts, boiler heat transfer, canister temperature distributions, steam plant size, and steam plant operation. Also, canister transportation is discussed since this influences operating costs. Details of economics of Army steam plants are provided including steam plant capital costs, operating costs, fuel reprocessor savings due to Army canister storage, and overall economics. Recommendations are made in the final section

  14. Review of advanced techniques for waste canister labeling

    International Nuclear Information System (INIS)

    Culbreth, W.G.; Bhagi, B.K.; Kanjerla, A.

    1994-01-01

    Technology has produced several new labeling techniques that may meet the needs of a nuclear waste repository. New methods must be capable of providing permanent labels to clearly identify the contents of each package containing high-level spent nuclear fuel. Several new techniques, along with their benefits and problems, are discussed

  15. Acceptance of spent nuclear fuel in multiple element sealed canisters by the Federal Waste Management System

    International Nuclear Information System (INIS)

    1990-03-01

    This report is one of a series of eight prepared by E.R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: (1) failed fuel; (2) consolidated fuel and associated structural parts; (3) non-fuel-assembly hardware; (4) fuel in metal storage casks; (5) fuel in multi-element sealed canisters; (6) inspection and testing requirements for wastes; (7) canister criteria; (8) spent fuel selection for delivery; and (9) defense and commercial high-level waste packages. 14 refs., 27 figs

  16. Initial results from the canistered waste forms produced during the first campaign of the DWPF Startup Test Program

    International Nuclear Information System (INIS)

    Harbour, J.R.

    1995-01-01

    As part of the Defense Waste Processing Facility (DWPF) Startup Test Program, approximately 90 canisters will be filled with glass containing simulated radioactive waste during five separate campaigns. The first campaign is a facility acceptance test to demonstrate the operability of the facility and to collect initial data on the glass and the canistered waste forms. During the next four campaigns (the waste qualification campaigns) data will be obtained which will be used to demonstrate that the DWPF product meets DOE's Waste Acceptance Product Specifications (WAPS). Currently 12 of the 16 canisters have been filled with glass during the first campaign (FA-13). This paper describes the tests that have been carried out on these 12 glass-filled canisters and presents the data with reference to the acceptance criteria of the WAPS. These tests include measurement of canister dimensions prior to and after glass filling. dew point, composition, and pressure of the gas within the free volume of the canister, fill height, free volume, weight, leak rates of welds and temporary seals, and weld parameters

  17. TOUGH: a numerical model for nonisothermal unsaturated flow to study waste canister heating effects

    International Nuclear Information System (INIS)

    Pruess, K.; Wang, J.S.Y.

    1983-12-01

    The physical processes modeled and the mathematical and numerical methods employed in a simulator for non-isothermal flow of water, vapor, and air in permeable media are briefly summarized. The simulator has been applied to study thermo-hydrological conditions in the near vicinity of high-level nuclear waste packages emplaced in unsaturated rocks. The studies reported here specifically address the question whether or not the waste canister environment will dry up in the thermal phase. 13 references, 8 figures, 2 tables

  18. Post-test evaluations of Waste Isolation Pilot Plant - Savannah River simulated defense HLW canisters and waste form

    International Nuclear Information System (INIS)

    Molecke, M.A.; Sorensen, N.R.; Harbour, J.R.; Ferrara, D.M.

    1993-01-01

    Eighteen nonradioactive defense high-level waste (DHLW) canisters were emplaced in and subjected to accelerated overtest thermal conditions for about three years at the bedded salt Waste Isolation Pilot Plant (WIPP) facility. Post-test laboratory corrosion results of several stainless steel 304L waste canisters, cast steel overpacks, and associated instruments ranged from negligible to moderate. We found appreciable surface corrosion and corrosion products on the cast steel overpacks. Pieces of both 304L and 316 stainless steel test apparatus underwent extensive stress-corrosion cracking failure and nonuniform attack. One of the retrieved test packages contained nonradioactive glass waste form from the Savannah River Site. We conducted post-test analyses of this glass to determine the degree of resultant glass fracturing, and whether any respirable fines were present. Linear glass fracture density ranged from about 1 to 8 fractures intersecting every 5 cm (2 inch) segment along a diameter line of the canister cross-section. Glass fines between 1 and 10 microns in diameter were detected, but were not quantified

  19. Genesis Solar Wind Science Canister Components Curated as Potential Solar Wind Collectors and Reference Contamination Sources

    Science.gov (United States)

    Allton, J. H.; Gonzalez, C. P.; Allums, K. K.

    2016-01-01

    The Genesis mission collected solar wind for 27 months at Earth-Sun L1 on both passive and active collectors carried inside of a Science Canister, which was cleaned and assembled in an ISO Class 4 cleanroom prior to launch. The primary passive collectors, 271 individual hexagons and 30 half-hexagons of semiconductor materials, are described in. Since the hard landing reduced the 301 passive collectors to many thousand smaller fragments, characterization and posting in the online catalog remains a work in progress, with about 19% of the total area characterized to date. Other passive collectors, surfaces of opportunity, have been added to the online catalog. For species needing to be concentrated for precise measurement (e.g. oxygen and nitrogen isotopes) an energy-independent parabolic ion mirror focused ions onto a 6.2 cm diameter target. The target materials, as recovered after landing, are described in. The online catalog of these solar wind collectors, a work in progress, can be found at: http://curator.jsc.nasa.gov/gencatalog/index.cfm This paper describes the next step, the cataloging of pieces of the Science Canister, which were surfaces exposed to the solar wind or component materials adjacent to solar wind collectors which may have contributed contamination.

  20. Acoustic monitoring techniques for corrosion degradation in cemented waste canisters

    International Nuclear Information System (INIS)

    Naish, C.C.; Buttle, D.; Wallace-Sims, R.; O'Brien, T.M.

    1991-01-01

    This report describes work to investigate acoustic emission as a non-intrusive monitor of corrosion and degradation of cemented wasteforms where the waste is a potentially reactive metal. The acoustic data collected shows good correlation with the corrosion rate as measured by hydrogen gas evolution rates and the electrochemically measured corrosion rates post cement hardening. The technique has been shown to be sensitive in detecting stress caused by expansive corrosion product within the cemented wasteform. The attenuation of the acoustic signal by the wasteform reduced the signal received by the monitoring equipment by a factor of 10 over a distance of approximately 150-400 mm, dependent on the water level in the cement. Full size packages were successfully monitored. It is concluded that the technique offers good potential for monitoring cemented containers of the more reactive metals, for example Magnox and aluminium. (author)

  1. Sensor-based automated docking of large waste canisters

    International Nuclear Information System (INIS)

    Drotning, W.D.

    1990-01-01

    Sensor-based programmable robots have the potential to speed up remote manipulation operations while protecting operators from exposure to radiation. Conventional master/slave manipulators have proven to be very slow in performing precision remote operations. In addition, inadvertent collisions of remotely manipulated objects with their environment increase the hazards associated with remote handling. This paper describes the development of a robotic system for the sensor-based automated remote manipulation and precision docking of large payloads. Computer vision and proximity sensing are used to control the precision docking of a large object with a passive target cavity. Specifically, a container of nuclear spent fuel on a transport vehicle is mated with an emplacement door on a vertical storage borehole at a waste repository

  2. 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

  3. Preliminary corrosion models for BWIP [Basalt Waste Isolation Project] canister materials

    International Nuclear Information System (INIS)

    Fish, R.L.; Anantatmula, R.P.

    1983-01-01

    Waste package development for the Basalt Waste Isolation Project (BWIP) requires the generation of materials degradation data under repository relevant conditions. These data are used to develop predictive models for the behavior of each component of waste package. The component models are exercised in performance analyses to optimize the waste package design. This document presents all repository relevant canister materials corrosion data that the BWIP and others have developed to date, describes the methodology used to develop preliminary corrosion models and provides the mathematical description of the models for both low carbon steel and Fe9Cr1Mo steel. Example environment/temperature history and model application calculations are presented to aid in understanding the models. The models are preliminary in nature and will be updated as additional corrosion data become available. 6 refs., 5 tabs

  4. Shielded Canister Transporter

    International Nuclear Information System (INIS)

    Eidem, G.G. Jr.; Fages, R.

    1993-01-01

    The Hanford Waste Vitrification Plant (HWVP) will produce canisters filled with high-level radioactive waste immobilized in borosilicate glass. This report discusses a Shielded Canister Transporter (SCT) which will provide the means for safe transportation and handling of the canisters from the Vitrification Building to the Canister Storage Building (CSB). The stainless steel canisters are 0.61 meters in diameter, 3.0 meters tall, and weigh approximately 2,135 kilograms, with a maximum exterior surface dose rate of 90,000 R/hr. The canisters are placed into storage tubes to a maximum of three tall (two for overpack canisters) with an impact limiter placed at the tube bottom and between each canister. A floor plug seals the top of the storage tube at the operating floor level of the CSB

  5. Corrosion of high-level radioactive waste iron-canisters in contact with bentonite.

    Science.gov (United States)

    Kaufhold, Stephan; Hassel, Achim Walter; Sanders, Daniel; Dohrmann, Reiner

    2015-03-21

    Several countries favor the encapsulation of high-level radioactive waste (HLRW) in iron or steel canisters surrounded by highly compacted bentonite. In the present study the corrosion of iron in contact with different bentonites was investigated. The corrosion product was a 1:1 Fe layer silicate already described in literature (sometimes referred to as berthierine). Seven exposition test series (60 °C, 5 months) showed slightly less corrosion for the Na-bentonites compared to the Ca-bentonites. Two independent exposition tests with iron pellets and 38 different bentonites clearly proved the role of the layer charge density of the swelling clay minerals (smectites). Bentonites with high charged smectites are less corrosive than bentonites dominated by low charged ones. The type of counterion is additionally important because it determines the density of the gel and hence the solid/liquid ratio at the contact to the canister. The present study proves that the integrity of the multibarrier-system is seriously affected by the choice of the bentonite buffer encasing the metal canisters in most of the concepts. In some tests the formation of a patina was observed consisting of Fe-silicate. Up to now it is not clear why and how the patina formed. It, however, may be relevant as a corrosion inhibitor. Copyright © 2014 Elsevier B.V. All rights reserved.

  6. Selection and evaluation of inner material candidates for Spanish high level radioactive waste canisters

    International Nuclear Information System (INIS)

    Puig, Francesc; Dies, Javier; Sevilla, Manuel; Pablo, Joan de; Pueyo, Juan Jose; Miralles, Lourdes; Martinez-Esparza, Aurora

    2007-01-01

    This paper summarizes the work carried out to analyse different alternatives related to the inner material selection of the Spanish high level waste canister for long term storage. The preliminary repository design considers granitic or clay formations, compacted bentonite sealing, corrosion allowing steel canisters and glass bead filling between the fuel assemblies and canister walls. This filling material will have the primary role of avoiding the possibility of a criticality event, which becomes an issue of major importance once the container is finally breached by corrosion and flooded by groundwater. In the first place, a complete set of requirements have been devised as evaluation criteria for candidate materials examination and selection; resulting in a compilation of demands significantly deeper and more exhaustive than any other similar work found in literature, including over 20 requirements and some other general aspects that could involve improvements in repository performance. Secondly, eight materials or material families (cast iron or steel, borosilicate glass, spinel, depleted uranium, dehydrated zeolites, hematite, phosphates and olivine) have been chosen and examined in detail, extracting some relevant conclusions. Either cast iron, borosilicate glass, spinel or depleted uranium are considered to look quite promising for the mentioned purpose. (authors)

  7. Selection of candidate canister materials for high-level nuclear waste containment in a tuff repository

    International Nuclear Information System (INIS)

    McCright, R.D.; Weiss, H.; Juhas, M.C.; Logan, R.W.

    1983-11-01

    A repository located at Yucca Mountain at the Nevada Test Site is a potential site for permanent geological disposal of high-level nuclear waste. The repository can be located in a horizon in welded tuff, a volcanic rock, which is above the static water level at this site. The environmental conditions in this unsaturated zone are expected to be air and water vapor dominated for much of the containment period. Type 304L stainless steel is the reference material for fabricating canisters to contain the solid high-level wastes. Alternative stainless alloys are considered because of possible susceptibility of 304L to localized and stress forms of corrosion. For the reprocessed glass wastes, the canisters serve as the recipient for pouring the glass with the result that a sensitized microstructure may develop because of the times at elevated temperatures. Corrosion testing of the reference and alternative materials has begun in tuff-conditioned water and steam environments. 21 references, 8 figures, 8 tables

  8. 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

  9. Modeling of thermal evolution of near field area around single pit mode nuclear waste canister disposal in soft rocks

    International Nuclear Information System (INIS)

    Bajpai, R.K.; Verma, A.K.; Maheshwar, Sachin

    2016-01-01

    Soft rocks like argillites/shales are under consideration worldwide as host rock for geological disposal of vitrified as well as spent fuel nuclear waste. The near field around disposed waste canister at 400-500m depth witnesses a complex heat field evolution due to varying thermal characteristics of rocks, coupling with hydraulic processes and varying intensity of heat flux from the canister. Smooth heat dissipation across the rock is desirable to avoid buildup of temperature beyond design limit (100 °C) and resultant micro fracturing due to thermal stresses in the rocks and intervening buffer clay layers. This also causes enhancement of hydraulic conductivity of the rocks, radionuclide transport and greater groundwater ingress towards the canister. Hence heat evolution modeling constitutes an important part of safety assessment of geological disposal facilities

  10. Critical review of welding technology for canisters for disposal of spent fuel and high level waste

    International Nuclear Information System (INIS)

    Pike, S.; Allen, C.; Punshon, C.; Threadgill, P.; Gallegillo, M.; Holmes, B.; Nicholas, J.

    2010-03-01

    Nagra is the Swiss national cooperative for the disposal of radioactive waste and is responsible for final disposal of all types of waste produced in Switzerland, which are partitioned into two repository types, one for spent fuel (SF), vitrified high-level waste (HLW) and long-lived intermediate level waste and one for low and intermediate level waste. In the general licences applied for these repositories, documentation has to show that long-term safety can be ensured and that factors for the construction, operation, and closure of the facility have been considered. Nagra has commissioned TWI to carry out a critical review of welding technologies for the sealing of HLW and SF canisters made of carbon steel. In conjunction with a material selection report, the information gained will be used as a preliminary step to provide input to developing design concepts for the canisters. The features to be considered are: a) Suitability of techniques for thickness of weld required; b) Suitability for remote operation, maintenance and set-up; c) Welding speed, weld quality, tolerances and cost; d) Effect of welding process on parent materials properties including microstructure corrosion resistance, distortion and residual stress; e) Potential post-weld treatments to reduce residual stress and enhance corrosion resistance; f) Suitability of inspection techniques for the weld thickness required; g) Impact of welding techniques on the canister design and material selection; h) Critique of emerging technologies which may be suitable in the future. The review of potential welding technologies began with a feasibility study carried out by TWI experts, where the unsuitable processes were rejected. For the remaining processes attention was focused on previous applications for the material and thickness suggested, and especially on safety critical applications such as applied in the nuclear and pressure vessel industry. Once the relevant information was gathered each process was

  11. The Swedish Concept for Disposal of Spent Nuclear Fuel: Differences Between Vertical and Horizontal Waste Canister Emplacement

    International Nuclear Information System (INIS)

    Bennett, D.G.; Hicks, T.W.

    2005-10-01

    The Swedish Nuclear Power Inspectorate (SKI) is preparing for the review of licence applications related to the disposal of spent nuclear fuel. The Swedish Nuclear Fuel and Waste Management Company (SKB) refers to its proposals for the disposal of spent nuclear fuel as the KBS-3 concept. In the KBS-3 concept, SKB plans that, after 30 to 40 years of interim storage, spent fuel will be disposed of at a depth of about 500 m in crystalline bedrock, surrounded by a system of engineered barriers. The principle barrier to radionuclide release is a cylindrical copper canister. Within the copper canister, the spent fuel is supported by a cast iron insert. Outside the copper canister is a layer of bentonite clay, known as the buffer, which is designed to provide mechanical protection for the canisters and to limit the access of groundwater and corrosive substances to their surfaces. The bentonite buffer is also designed to sorb radionuclides released from the canisters, and to filter any colloids that may form within the waste. SKB is expected to base its forthcoming licence applications on a repository design in which the waste canisters are emplaced in vertical boreholes (KBS-3V). However, SKB has also indicated that it might be possible and, in some respects, beneficial to dispose of the waste canisters in horizontal tunnels (KBS-3H). There are many similarities between the KBS-3V and KBS-3H designs. There are, however, uncertainties associated with both of the designs and, when compared, both possess relative advantages and disadvantages. SKB has identified many of the key factors that will determine the evolution of a KBS-3H repository and has plans for research and development work in many of the areas where the differences between the KBS-3V and KBS-3H designs mean that they could be significant in terms of repository performance. With respect to the KBS-3H design, key technical issues are associated with: 1. The accuracy of deposition drift construction. 2. Water

  12. Preliminary assessment of the thermal effects of an annular air space surrounding an emplaced nuclear waste canister

    International Nuclear Information System (INIS)

    Davis, B.W.

    1979-01-01

    Modeling results have previously shown that the presence of a large air space (e.g., a repository room) within a nuclear waste repository is expected to cause a waste canister's temperature to remain cooler than it would otherwise be. Results presented herein show that an annular air space surrounding the waste canisters can have similar cooling effects under certain prescribable conditions; for a 16 ft x 1 ft diameter canister containing 650 PWR rods which initially generate a total of 4.61 kw, analysis will show that annular air spaces greater than 11 in will permit the canister surface to attain peak temperatures lower than that which would result from a zero-gap/perfect thermal contact. It was determined that the peak radial temperature gradient in the salt varies in proportion to the inverse of the drill hole radius. Thermal radiation is shown to be the dominant mode of heat transfer across an annular air space during the first two years after emplacement. Finally, a methodology is presented which will allow investigators to easily model radiation and convection heat transfer through air spaces by treating the space as a conduction element that possesses non-linear temperature dependent conductivity

  13. Rates and mechanisms of radioactive release and retention inside a waste disposal canister - in Can Processes

    Energy Technology Data Exchange (ETDEWEB)

    Oversby, V.M. (ed.) [and others

    2003-10-01

    the system that will not be present under long term disposal conditions. A simulation of long-term conditions can be done using uranium dioxide that contains a short-lived isotope of uranium, but this will not include the effects of fission product and higher actinide elements on the behaviour of the spent fuel. We designed a project that had as its objective to improve the scientific understanding of the processes that control release of radioactive species from spent fuel inside a disposal canister and the chemical changes in those species that might limit release of radioactivity from the canister. If the mechanisms that control dissolution of the fuel matrix, including self-irradiation effects, can be clarified, a more realistic assessment of the long-term behaviour of spent fuel under disposal conditions can be made. By removing uncertainties concerning waste form performance, a better assessment of the individual and collective role of the engineered barriers can be made. To achieve the overall objective of the project, the following scientific and technical objectives were set. 1. Measure the actual rate of matrix dissolution of uranium dioxide under oxidising and reducing conditions. 2. Measure the effect of alpha radiolysis on the dissolution rate of uranium dioxide under oxidising and reducing conditions. 3. Measure the dissolution rate of the matrix material of spent fuel and thereby determine the additional effects of beta and gamma radiation on uranium dioxide dissolution rate under oxidising and reducing conditions. 4. Measure the ability of actively corroding iron to reduce oxidised U(VI) to U(IV) when U is present as the complex ion uranyl carbonate. 5. Measure the rate of reduction of Np(V) species in the presence of actively corroding iron. 6. Calculate the expected equilibrium and steady state concentrations of U under the conditions of the experiments used for meeting objectives 1 through 3 and compare the calculated results with those measured in

  14. Corrosion of high-level radioactive waste iron-canisters in contact with bentonite

    Energy Technology Data Exchange (ETDEWEB)

    Kaufhold, Stephan, E-mail: s.kaufhold@bgr.de [BGR, Bundesanstalt für Geowissenschaften und Rohstoffe, Stilleweg 2, D-30655 Hannover (Germany); Hassel, Achim Walter [Max-Planck-Institut für Eisenforschung GmbH, Max-Planck-Straße 1, D-40237 Düsseldorf (Germany); Institute for Chemical Technology of Inorganic Materials, Johannes Kepler University Linz, Altenberger Straße 69, 4040 Linz (Austria); Sanders, Daniel [Max-Planck-Institut für Eisenforschung GmbH, Max-Planck-Straße 1, D-40237 Düsseldorf (Germany); Dohrmann, Reiner [BGR, Bundesanstalt für Geowissenschaften und Rohstoffe, Stilleweg 2, D-30655 Hannover (Germany); LBEG, Landesamt für Bergbau, Energie und Geologie, Stilleweg 2, D-30655 Hannover (Germany)

    2015-03-21

    Graphical abstract: Corrosion at the bentonite iron interface proceeds unaerobically with formation of an 1:1 Fe silicate mineral. A series of exposure tests with different types of bentonites showed that Na–bentonites are slightly less corrosive than Ca–bentonites and highly charges smectites are less corrosive compared to low charged ones. The formation of a patina was observed in some cases and has to be investigated further. - Highlights: • At the iron bentonite interface a 1:1 Fe layer silicate forms upon corrosion. • A series of iron–bentonite corrosion products showed slightly less corrosion for Na-rich and high-charged bentonites. • In some tests the formation of a patina was observed consisting of Fe–silicate, which has to be investigated further. - Abstract: Several countries favor the encapsulation of high-level radioactive waste (HLRW) in iron or steel canisters surrounded by highly compacted bentonite. In the present study the corrosion of iron in contact with different bentonites was investigated. The corrosion product was a 1:1 Fe layer silicate already described in literature (sometimes referred to as berthierine). Seven exposition test series (60 °C, 5 months) showed slightly less corrosion for the Na–bentonites compared to the Ca–bentonites. Two independent exposition tests with iron pellets and 38 different bentonites clearly proved the role of the layer charge density of the swelling clay minerals (smectites). Bentonites with high charged smectites are less corrosive than bentonites dominated by low charged ones. The type of counterion is additionally important because it determines the density of the gel and hence the solid/liquid ratio at the contact to the canister. The present study proves that the integrity of the multibarrier-system is seriously affected by the choice of the bentonite buffer encasing the metal canisters in most of the concepts. In some tests the formation of a patina was observed consisting of Fe

  15. SRL canister impact tests

    International Nuclear Information System (INIS)

    Kelker, J.W. Jr.

    1986-05-01

    The Defense Waste Processing Facility (DWPF) is being constructed at the SRP for the containerization of high-level nuclear waste as a waste form for eventual permanent disposal. The waste will be incorporated in molten glass and solidified in Type 304L stainless steel canisters 2 feet in diameter x 9 feet 10 inches long. The canisters have a minimum wall thickness of 3/8 inch. Over a three-year period, nineteen drop-tests of nine canisters, filled with simulated waste glass, were made in support of the DWPF containerization program. Eight of the canister evaluation tests were of Type 304L stainless steel material and one was of commercially pure titanium. Three different length (9.44, 5.06, and 7.88 inch) nozzle configurations containing final closure upset welds were evaluated for the stainless steel canisters. All impact tests of the stainless steel canisters, which included bottom-, side-, and top-drops, were acceptable. The bottom-drop test of the titanium canister, which contained a final closure upset weld, was acceptable; however, the top-drop resulted in a breaching of the top head where it joins the nozzle. The final closure titanium upset weld was acceptable. The titanium canister wall thickness was 1/4 inch

  16. Canister materials proposed for final disposal of high level nuclear waste - a review with respect to corrosion resistance

    Energy Technology Data Exchange (ETDEWEB)

    Mattsson, E; Odoj, R; Merz, E [eds.

    1981-06-01

    Spent fuel from nuclear reactors has to be disposed of either after reprocessing or without such treatment. Due to toxic radiation the nuclear waste has to be isolated from the biosphere for 300-1000 years, or in extreme cases for more than 100,000 years. The nuclear waste will be enclosed in corrosion resistant canisters. These will be deposited in repositories in geological formations, such as granite, basalt, clay, bedded or domed salt, or the sediments beneath the deep ocean floor. There the canisters will be exposed to groundwater, brine or seawater at an elevated temperature. Species formed by radiolysis may affect the corrosivity of the agent. The corrosion resistance of candidate canister materials is evaluated by corrosion tests and by thermodynamic and mass transport calculations. Examination of ancient metal objects after long exposure in nature may give additional information. On the basis of the work carried out so far, the principal candidate canister materials are titanium materials, copper and high purity alumina.

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  18. Production methods and costs of oxygen free copper canisters for nuclear waste disposal

    International Nuclear Information System (INIS)

    Rajainmaeki, H.; Nieminen, M.; Laakso, L.

    1991-08-01

    The fabrication technology and costs of various manufacturing alternatives to make large copper canisters for spent fuel repository are discussed. The capsule design is based on the TVO's new advanced cold process concept where a steel canister is surrounded by the oxygen free copper canister. This study shows that already at present there exist several possible manufacturing routes, which result in consistently high quality canisters. Hot rolling, bending and EB-welding the seam is the best way to assure the small grain size which is preferable for the best inspectability of the final EB-welded seam of the lid. The same route turns out also to be the most economical

  19. Production methods and costs of oxygen free copper canisters for nuclear waste disposal

    International Nuclear Information System (INIS)

    Rajainmaeki, H.; Nieminen, M.; Laakso, L.

    1991-06-01

    The fabrication technology and costs of various manufacturing alternatives to make large copper canisters for spent fuel repository are discussed. The capsule design is based on the TVO's new advanced cold process concept where a steel canister is surrounded by the oxygen free copper canister. This study shows that already at present there exist several possible manufacturing routes, which results in consistently high quality canisters. Hot rolling, bending and EB-welding the seam is the best way to assure the small grain size which is preferable for the best inspectability of the final EB-welded seam of the lid. The same route turns out also to be the most economical. (au)

  20. Brine: a computer program to compute brine migration adjacent to a nuclear waste canister in a salt repository

    International Nuclear Information System (INIS)

    Duckworth, G.D.; Fuller, M.E.

    1980-01-01

    This report presents a mathematical model used to predict brine migration toward a nuclear waste canister in a bedded salt repository. The mathematical model is implemented in a computer program called BRINE. The program is written in FORTRAN and executes in the batch mode on a CDC 7600. A description of the program input requirements and output available is included. Samples of input and output are given

  1. Retrieval of canistered experimental waste at the waste isolation pilot plant

    International Nuclear Information System (INIS)

    Stinebaugh, R.E.

    1979-07-01

    To assess the suitability of bedded salt for nuclear waste disposal, an extensive experimental program will be implemented at the Waste Isolation Pilot Plant. In order to evaluate experimental results, it will be necessary to recover certain of these experiments for postmortem examination and analysis. This document describes the equipment and procedures used to effect recovery of one category of WIPP experiments

  2. Corrosion research of alloys for use in nuclear waste disposal canisters

    International Nuclear Information System (INIS)

    Sorensen, N.R.; Diegle, R.B.

    1984-01-01

    All tests involving the corrosion behavior of TiCode-12 (Ti-0.3Mo-0.8Ni) indicate that this alloy is a suitable choice for the containment of nuclear wastes. Although cracking was observed under conditions of hydrogen embrittlement, and pitting occurred in aggressive media, the conditions which produce these two failure modes represented severe overtests and are not expected in a repository. Both Mo and Ni increase the degree of corrosion resistance of Ti alloys. Mo changes the nature of the oxide, while Ni forms intermetallic precipitates which, through galvanic coupling, polarize the Ti into the passive region. All alternate alloys tested (including cast iron) exhibit acceptable corrosion behavior relative to their role as backup canister materials. In the case of cast iron, the material provides for a corrosion allowance design in contrast to the corrosion resistant design offered by Ti alloys. The effects of gamma irradiation still need to be adequately characterized. Research on such effects is underway at Sandia and the University of Minnesota. 11 references, 5 figures, 1 table

  3. Development of the DWPF canister temporary shrink-fit seal

    International Nuclear Information System (INIS)

    Kelker, J.W. Jr.

    1986-04-01

    The Defense Waste Processing Facility is being constructed at The Savannah River Plant for the containerization of high-level nuclear waste in a wasteform for eventual permanent disposal. The waste will be incorporated in molten glass and solidified in type 304L stainless steel canisters, 2-feet in diameter x 9-feet 10-inches long, containing a flanged 6-in.-diam pipe fill-nozzle. The canisters have a minimum wall thickness of 3/8 in. Utilizing the heat from the glass filling operation, a shrink-fit seal for a plug in the end of the canister fill nozzle was developed that: will withstand the radioactive environment; will prevent the spread of contamination, and will keep moisture and water from entering the canister during storage and decontamination of the canister by wet-frit blasting to remove smearable and oxide-film fixed radioactive nuclides; is removable and can be replaced by a new oversize plug in the event the seal fails the pressure decay leakage test ( -4 atm cc/sec helium); will keep the final weld closure clean and free of nuclear contamination; will withstand being pressed into the nozzle without exposing external contamination or completely breaking the seal; is reliable; and is easily installed. The seal consists of: a removable sleeve (with a tapered bore) which is shrink-fitted into the nozzle bore during canister fabrication; and a tapered plug which is placed into the sleeved nozzle after the canister is filled with radioactive molten glass. A leak-tight shrink-fit seal is formed between the nozzle, sleeve, and plug upon temperature equilibrium. The temporarily sealed canister is transferred from the Melt cell to the Decon cell, and the surface is decontaminated. Next it is transferred to the Weld/Test cell where the temporary seal is pressed down into the nozzle, revealing a clean cavity where the canister final closure weld is made

  4. Production methods and costs of oxygen free copper canisters for nuclear waste disposal

    International Nuclear Information System (INIS)

    Aalto, H.; Rajainmaeki, H.; Laakso, L.

    1996-10-01

    The fabrication technology and costs of various manufacturing alternatives to make large copper canisters for disposal of spent nuclear fuel from reactors of Teollisuuden Voima Oy (TVO) and Imatran Voima Oy (IVO) are discussed. The canister design is based on the Posiva's concept where solid insert structure is surrounded by the copper mantle. During recent years Outokumpu Copper Products and Posiva have continued their work on development of the copper canisters. Outokumpu Copper Products has also increased capability to manufacture these canisters. In the study the most potential manufacturing methods and their costs are discussed. The cost estimates are based on the assumption that Outokumpu will supply complete copper mantles. At the moment there are at least two commercially available production methods for copper cylinder manufacturing. These routes are based on either hot extrusion of the copper tube or hot rolling, bending and EB-welding of the tube. Trial fabrications has been carried out with both methods for the full size canisters. These trials of the canisters has shown that both the forming from rolled plate and the extrusion are possible methods for fabricating copper canisters on a full scale. (orig.) (26 refs.)

  5. The role of the canister in a system for the final disposal of spent fuel or high-level waste

    International Nuclear Information System (INIS)

    Papp, T.

    1986-01-01

    A final repository for radioactive waste must isolate the toxic substances or distribute their release over time or space to avoid causing harmful concentrations of radionuclides in the biosphere. The Swedish research has focused on a repository 500 m down in crystalline rock where the geochemical environment can give canisters a service life of the order of a million years. These evaluations are discussed and the safety effect of the canister is compared with that of other barriers available in a repository system. Our conclusions are that a combined protection effect of natural and man-made barriers can be achieved that substantially exceeds what could reasonably be required by society. An actual repository design can then be based on an optimization of the cost to reach a level of accepted safety with due regard for the safety margins and redundancy necessary for achieving public confidence. (author)

  6. Melting of contaminated metallic waste

    International Nuclear Information System (INIS)

    Lee, Y.-S.; Cheng, S.-Y.; Kung, H.-T.; Lin, L.-F.

    2004-01-01

    Approximately 100 tons of contaminated metallic wastes were produced each year due to maintenance for each TPC's nuclear power reactor and it was roughly estimated that there will be 10,000 tons of metallic scraps resulted from decommissioning of each reactor in the future. One means of handling the contaminated metal is to melt it. Melting process owns not only volume reduction which saves the high cost of final disposal but also resource conservation and recycling benefits. Melting contaminated copper and aluminum scraps in the laboratory scale have been conducted at INER. A total of 546 kg copper condenser tubes with a specific activity of about 2.7 Bq/g was melted in a vacuum induction melting facility. Three types of products, ingot, slag and dust were derived from the melting process, with average activities of 0.10 Bq/g, 2.33 Bq/g and 84.3 Bq/g respectively. After the laboratory melting stage, a pilot plant with a 500 kg induction furnace is being designed to melt the increasingly produced contaminated metallic scraps from nuclear facilities and to investigate the behavior of different radionuclides during melting. (author)

  7. CANISTER TRANSFER SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    B. Gorpani

    2000-01-01

    The Canister Transfer System receives transportation casks containing large and small disposable canisters, unloads the canisters from the casks, stores the canisters as required, loads them into disposal containers (DCs), and prepares the empty casks for re-shipment. Cask unloading begins with cask inspection, sampling, and lid bolt removal operations. The cask lids are removed and the canisters are unloaded. Small canisters are loaded directly into a DC, or are stored until enough canisters are available to fill a DC. Large canisters are loaded directly into a DC. Transportation casks and related components are decontaminated as required, and empty casks are prepared for re-shipment. One independent, remotely operated canister transfer line is provided in the Waste Handling Building System. The canister transfer line consists of a Cask Transport System, Cask Preparation System, Canister Handling System, Disposal Container Transport System, an off-normal canister handling cell with a transfer tunnel connecting the two cells, and Control and Tracking System. The Canister Transfer System operating sequence begins with moving transportation casks to the cask preparation area with the Cask Transport System. The Cask Preparation System prepares the cask for unloading and consists of cask preparation manipulator, cask inspection and sampling equipment, and decontamination equipment. The Canister Handling System unloads the canister(s) and places them into a DC. Handling equipment consists of a bridge crane hoist,; DC--loading manipulator, lifting fixtures, and small canister staging racks. Once the cask has been unloaded, the Cask Preparation System decontaminates the cask exterior and returns it to the Carrier/Cask Handling System via the Cask Transport System. After the; DC--is fully loaded, the Disposal Container Transport System moves the; DC--to the Disposal Container Handling System for welding. To handle off-normal canisters, a separate off-normal canister

  8. Canister Transfer System Description Document

    International Nuclear Information System (INIS)

    2000-01-01

    The Canister Transfer System receives transportation casks containing large and small disposable canisters, unloads the canisters from the casks, stores the canisters as required, loads them into disposal containers (DCs), and prepares the empty casks for re-shipment. Cask unloading begins with cask inspection, sampling, and lid bolt removal operations. The cask lids are removed and the canisters are unloaded. Small canisters are loaded directly into a DC, or are stored until enough canisters are available to fill a DC. Large canisters are loaded directly into a DC. Transportation casks and related components are decontaminated as required, and empty casks are prepared for re-shipment. One independent, remotely operated canister transfer line is provided in the Waste Handling Building System. The canister transfer line consists of a Cask Transport System, Cask Preparation System, Canister Handling System, Disposal Container Transport System, an off-normal canister handling cell with a transfer tunnel connecting the two cells, and Control and Tracking System. The Canister Transfer System operating sequence begins with moving transportation casks to the cask preparation area with the Cask Transport System. The Cask Preparation System prepares the cask for unloading and consists of cask preparation manipulator, cask inspection and sampling equipment, and decontamination equipment. The Canister Handling System unloads the canister(s) and places them into a DC. Handling equipment consists of a bridge crane/hoist, DC loading manipulator, lifting fixtures, and small canister staging racks. Once the cask has been unloaded, the Cask Preparation System decontaminates the cask exterior and returns it to the Carrier/Cask Handling System via the Cask Transport System. After the DC is fully loaded, the Disposal Container Transport System moves the DC to the Disposal Container Handling System for welding. To handle off-normal canisters, a separate off-normal canister handling

  9. Development and demonstration of prototype transportation equipment for emplacing HL vitrified waste canisters into small diameter bored horizontal disposal cells

    International Nuclear Information System (INIS)

    Seidler, Wolf K.; Bosgiraud, Jean-Michel; Londe, Louis

    2008-01-01

    Over a period of 4 and years the National Radioactive Waste Management Agency (Andra), working with a variety of Contractors mostly specializing in nuclear orientated mechanical applications, successfully designed, fabricated and demonstrated 2 very different prototype high level waste transport systems. The first system, based on air cushion technology, was developed primarily for very heavy loads (17 to 45 tonnes). The results of this work are described in a separate presentation (Paper 21) at this Conference. The second system, developed by Andra within the framework of the ESDRED Project, generally referred to as the 'Pushing Robot System' for vitrified waste canisters, is the subject of this paper. The 'Pushing Robot System' is a part of the French national disposal concept that is described in Andra's 'Dossier 2005'. The latter is a public document that can be viewed on Andra's web site (www.andra.fr). The 'Pushing Robot System' system is designed for the deep geological disposal (in clay formations) of 'C' type vitrified waste canisters. In its entirety the system provides for the transport, emplacement and, if necessary, the retrieval of those canisters. Nothing in the design of the Andra emplacement equipment would preclude its utilization in horizontal openings in other types of geological settings. Over a period of some 8 years Andra has developed the 'Pushing Robot System' in 3 phases. Initially there was only the 'Conceptual Design' (Phase 1) which was incorporated in the Dossier 2005. This was followed by Phase 2 i.e. the design and fabrication of a simplified full scale prototype system henceforth referred to a P1, which includes a Pushing Robot, a Dummy Canister and a Test Bench. P1 details were also incorporated in the Dossier 2005. Finally, during Phase 3, a second more comprehensive full scale prototype system P2 has been designed and is being assembled and tested this month. This system includes a Transport Shuttle, a Transfer Shielding Cask, a

  10. Monitoring of plutonium contaminated solid waste streams

    International Nuclear Information System (INIS)

    Birkhoff, G.; Notea, A.

    1977-01-01

    The planning of a system for monitoring Pu contaminated solid waste streams, from the nuclear fuel cycle, is considered on the basis of given facility waste management program. The inter relations between the monitoring system and the waste management objectives are stressed. Selection criteria with pertinent data of available waste monitors are given. Example of monitoring systems planning are presented and discussed

  11. Computer Modeling Of High-Level Waste Glass Temperatures Within DWPF Canisters During Pouring And Cool Down

    International Nuclear Information System (INIS)

    Amoroso, J.

    2011-01-01

    This report describes the results of a computer simulation study to predict the temperature of the glass at any location inside a DWPF canister during pouring and subsequent cooling. These simulations are an integral part of a larger research focus aimed at developing methods to predict, evaluate, and ultimately suppress nepheline formation in HLW glasses. That larger research focus is centered on holistically understanding nepheline formation in HLW glass by exploring the fundamental thermal and chemical driving forces for nepheline crystallization with respect to realistic processing conditions. Through experimental work, the goal is to integrate nepheline crystallization potential in HLW glass with processing capability to ultimately optimize waste loading and throughput while maintaining an acceptable product with respect to durability. The results of this study indicated severe temperature gradients and prolonged temperature dwell times exist throughout different locations in the canister and that the time and temperatures that HLW glass is subjected to during processing is a function of pour rate. The simulations indicate that crystallization driving forces are not uniform throughout the glass volume in a DWPF (or DWPF-like) canister and illustrate the importance of considering overall kinetics (chemical and thermal driving forces) of nepheline formation when developing methods to predict and suppress its formation in HLW glasses. The intended path forward is to use the simulation data both as a driver for future experimental work and, as an investigative tool for evaluating the impact of experimental results. Simulation data will be used to develop laboratory experiments to more acutely evaluate nepheline formation in HLW glass by incorporating the simulated temperatures throughout the canister into the laboratory experiments. Concurrently, laboratory experiments will be performed to identify nepheline crystallization potential in HLW glass as a function of

  12. HLW Canister and Can-In-Canister Drop Calculation

    International Nuclear Information System (INIS)

    H. Marr

    1999-01-01

    The purpose of this calculation is to evaluate the structural response of the standard high-level waste (HLW) canister and the HLW canister containing the cans of immobilized plutonium (''can-in-canister'' throughout this document) to the drop event during the handling operation. The objective of the calculation is to provide the structure parameter information to support the canister design and the waste handling facility design. Finite element solution is performed using the commercially available ANSYS Version (V) 5.4 finite element code. Two-dimensional (2-D) axisymmetric and three-dimensional (3-D) finite element representations for the standard HLW canister and the can-in-canister are developed and analyzed using the dynamic solver

  13. A review of materials and corrosion issues regarding canisters for disposal of spent fuel and high-level waste in Opalinus clay

    International Nuclear Information System (INIS)

    Landolt, D.; Davenport, A.; Payer, J.; Shoesmith, D.

    2009-01-01

    The project 'Entsorgungsnachweis' presented by NAGRA to the Swiss Federal Government in December 2002 assessed the feasibility of disposal of spent fuel (SF), vitrified high level waste (HLW) from reprocessing and long-lived intermediate level waste in an Opalinus Clay repository site in Northern Switzerland. NAGRA proposed the use of carbon steel canisters for disposal of SF/HLW and it also put forward an alternative concept of copper canisters with cast iron insert. In its reply the Federal Government acknowledged that NAGRA had successfully demonstrated the technical feasibility of disposal of SF/HLW. However, some of its experts raised a number of questions related to the choice of steel as canister material. Among others, it was questioned whether hydrogen formed by corrosion of steel in contact with saturated bentonite might adversely affect the barrier function of the Opalinus clay. It was also recommended that alternative canister materials and/or design concepts should be evaluated. To deal with these concerns NAGRA convened an international group of experts, the Canister Materials Review Board (CMRB), who were to review the existing information on canister materials that could be suitable for the proposed repository environment. Based on present knowledge of materials science, the CMRB was to recommend to NAGRA the most suitable material(s) for meeting the performance requirements for SF/HLW canisters. Specifically, the CMRB was to consider corrosion, including hydrogen generation, and stress-assisted failure processes that could affect the integrity and projected life time of SF/HLW canisters or impede the functioning of geological barriers while keeping in mind the overall feasibility of manufacturing, sealing and inspecting the canisters. The CMRB was further asked to identify the needs and provide advice for further studies by NAGRA on the long term performance and safety of SF/HLW canisters in the Swiss repository concept. For the assessment of the

  14. Neutron interrogator assay system for the Idaho Chemical Processing Plant waste canisters and spent fuel: preliminary description and operating procedures manual

    International Nuclear Information System (INIS)

    Menlove, H.O.; Eccleston, G.; Close, D.A.; Speir, L.G.

    1978-05-01

    A neutron interrogation assay system is being designed for the measurement of waste canisters and spent fuel packages at the new Idaho Chemical Processing Plant to be operated by Allied Chemical Corp. The assay samples consist of both waste canisters from the fluorinel dissolution process and spent fuel assemblies. The assay system is a 252 Cf ''Shuffler'' that employs a cyclic sequence of fast-neutron interrogation with a 252 Cf source followed by delayed-neutron counting to determine the 235 U content

  15. Safeguards considerations related to the use of multi-purpose canisters in the Civilian Radioactive Waste Management system

    International Nuclear Information System (INIS)

    Floyd, W.C.

    1995-01-01

    The US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is responsible for disposing of the nation's high-level radioactive waste. Currently, DOE is considering the use of Multi-Purpose Canisters (MPCs) to containerize commercial spent nuclear fuel (SNF) to be handled by the system. To achieve its safeguards and security objectives, OCRWM plans to institute a US Regulatory Commission (NRC)-approved safeguards program. Since the Mined Geologic Disposal System (MGDS) facility and a possible Monitored Retrievable Storage (MRS) facility may be subject to selection for International Atomic Energy Agency (IAEA) inspections, the safeguards program for MPCs may not preclude compliance with the requirements of the IAEA's Annex D, Special Criteria for Difficult-to-Access Fuel Items. MPC safeguards are based on three principles: Verification, Material Control and Accounting, and Physical Protection

  16. The suitability of Titanium as a corrosion resistant canister for nuclear waste

    International Nuclear Information System (INIS)

    Henriksson, S.; Pettersson, K.

    1977-08-01

    A literature study and inventory of experience has been carried out, aimed at assessing the possibilities of unalloyed and Pd-alloyed titanium withstanding corrosion for 1000 - 10000 years in contact with Baltic Sea water at 100 percentC and pH 4 - 10. The following assesment can be made: -1. Pitting, crevice corrosion, stress corrosion cracking and corrosion fatigue constitute no problem if the canister is made of unalloyed titanium corresponding to ASTM Grade 1. 2. Linear extrapolation of reported corrosion rates for oxidation and general corrosion gives a life of between 1000 and 10000 years for a 5 mm thick canister. 3. Hydrogen embrittlement resulting from hydrogen pick-up from the deposition environment should not occur. Delayed failure caused by a redistribution of the hydrogen initially present in the titanium can be avoided if its concentration is maximized to 20 ppm. (author)

  17. Feasibility of using a high-level waste canister as an engineered barrier in disposal

    International Nuclear Information System (INIS)

    Slate, S.C.; Pitman, S.G.; Nesbitt, J.F.; Partain, W.L.

    1982-08-01

    The objective of this report is to evaluate the feasibility of designing a process canister that could also serve as a barrier canister. To do this a general set of performance criteria is assumed and several metal alloys having a high probability of demonstrating high corrosion resistance under repository conditions are evaluated in a qualitative design assessment. This assessment encompasses canister manufacture, the glass-filling process, interim storage, transportation, and to a limited extent, disposal in a repository. A series of scoping tests were carried out on two titanium alloys and Inconel 625 to determine if the high temperature inherent in the glass-fill processing would seriously affect either the strength or corrosion resistance of these metals. This is a process-related concern unique to the barrier canister concept. The material properties were affected by the heat treatments which simulated both the joule-heated glass melter process (titanium alloys and Inconel 625) and the in-can melter (ICM) process (Inconel 625). However, changes in the material properties were generally within 20% of the original specimens. Accelerated corrosion testing of the heat treated coupons in a highly oxygenated brine showed basic corrosion resistance of titanium grade 12 and Inconel 625 to compare favorably with that of the untreated coupons. The titanium grade 2 coupons experienced severe corrosion pitting. These corrosion tests were of a scoping nature and suitable primarily for the detection of gross sensitivity to the heat treatment inherent in the glass-fill process. They are only suggstive of repository performance since the tests do not adequately model the wide range of repository conditions that could conceivably occur

  18. Preparing, Loading and Shipping Irradiated Metals in Canisters Classified as Remote-Handled (RH) Low-Level Waste (LLW) From Oak Ridge National Laboratory (ORNL) to the Nevada Test Site (NTS)

    International Nuclear Information System (INIS)

    McClelland, B.C.; Moore, T.D.

    2006-01-01

    Irradiated metals, classified as remote-handled low-level waste generated at the Oak Ridge National Laboratory (ORNL) in Oak Ridge, Tennessee, were containerised in various sized canisters for long-term storage. The legacy waste canisters were placed in below-grade wells located at the 7827 Facility until a pathway for final disposal at the Nevada Test Site (NTS) could be identified and approved. Once the pathway was approved, WESKEM, LLC was selected by Bechtel Jacobs Company, LLC to prepare, load, and ship these canisters from ORNL to the NTS. This paper details some of the technical challenges encountered during the retrieval process and solutions implemented to ensure the waste was safely and efficiently over-packed and shipped for final disposal. The technical challenges detailed in this paper include: 1) how to best perform canister/lanyard pre-lift inspections since some canisters had not been moved in ∼10 years, so deterioration was a concern; 2) replacing or removing damaged canister lanyards; 3) correcting a mis-cut waste canister lanyard resulting in a shielded overpack lid not seating properly; 4) retrieving a stuck canister; and 5) developing a path forward after an overstrained lanyard failed causing a well shield plug to fall and come in contact with a waste canister. Several of these methods can serve as positive lessons learned for other projects encountering similar situations. (authors)

  19. Analysis of heat and mass transport processes near an emplaced nuclear waste canister

    International Nuclear Information System (INIS)

    Keller, C.

    1990-01-01

    A review has been performed of the models and experimental plans for evaluation of the spent fuel canister environment in a nuclear repository, e.g., the planned Yucca Mountain facilities. Special emphasis was placed on the relevance of the models and experiments to the 100 to 10,000 year prediction. The question was addressed whether one could justify testing in materials other than Yucca Mountain rock and obtain results in a relatively short time which would be relevant to the long time in Yucca Mountain. The paper discusses steam evolution in calculations and experiments, fracture models, possible measurements of relative permeability, and long time scale effects. 5 figs. (MB)

  20. Waste management of actinide contaminated soil

    International Nuclear Information System (INIS)

    Navratil, J.D.; Thompson, G.H.; Kochen, R.L.

    1978-01-01

    Waste management processes have been developed to reduce the volume of Rocky Flats soil contaminated with plutonium and americium and to prepare the contaminated fraction for terminal storage. The primary process consists of wet-screening. The secondary process uses attrition scrubbing and wet screening with additives. The tertiary process involves volume reduction of the contaminated fraction by calcination, or fixation by conversion to glass. The results of laboratory scale testing of the processes are described

  1. Investigation into the suitability of titanium as a corrosion resistant canister for nuclear waste

    International Nuclear Information System (INIS)

    Henriksson, S.; Pettersson, J.

    A literature study and inventory of experience has been carried out, aimed at assessing the possibilities of unalloyed and Pd-alloyed titanium withstanding corrosion for 1,000 to 10,000 years in contact with Baltic Sea water at 100 0 C and pH 4 to 10. Pitting, crevice corrosion, stress corrosion cracking and corrosion fatigue constitute no problem if the canister is made of unalloyed titanium corresponding to ASTM Grade 1. Titanium alloyed with palladium therefore need not be used. Linear extrapolation of reported corrosion rates for oxidation and general corrosion gives a life of between 1,000 and 10,000 years for a 5 mm thick canister. This estimate must be considered to be conservative since oxidation in fact follows a logarithmic law. Hydrogen embrittlement resulting from hydrogen pick-up from the deposition environment should not occur. Delayed failure caused by a redistribution of the hydrogen initially present in the titanium can be avoided if its concentration is maximized to 20 ppM. Pd-alloyed titanium is more sensitive than unalloyed titanium to hydrogen pick-up, especially in galvanic contact with less noble metals

  2. TRADITIONAL CANISTER-BASED OPEN WASTE MANAGEMENT SYSTEM VERSUS CLOSED SYSTEM: HAZARDOUS EXPOSURE PREVENTION AND OPERATING THEATRE STAFF SATISFACTION.

    Science.gov (United States)

    Horn, M; Patel, N; MacLellan, D M; Millard, N

    2016-06-01

    Exposure to blood and body fluids is a major concern to health care professionals working in operating rooms (ORs). Thus, it is essential that hospitals use fluid waste management systems that minimise risk to staff, while maximising efficiency. The current study compared the utility of a 'closed' system with a traditional canister-based 'open' system in the OR in a private hospital setting. A total of 30 arthroscopy, urology, and orthopaedic cases were observed. The closed system was used in five, four, and six cases, respectively and the open system was used in nine, two, and four cases, respectively. The average number of opportunities for staff to be exposed to hazardous fluids were fewer for the closed system when compared to the open during arthroscopy and urology procedures. The open system required nearly 3.5 times as much staff time for set-up, maintenance during procedures, and post-procedure disposal of waste. Theatre staff expressed greater satisfaction with the closed system than with the open. In conclusion, compared with the open system, the closed system offers a less hazardous and more efficient method of disposing of fluid waste generated in the OR.

  3. Geological Disposal of Nuclear Waste: Investigating the Thermo-Hygro-Mechanical-Chemical (THMC) Coupled Processes at the Waste Canister- Bentonite Barrier Interface

    Science.gov (United States)

    Davies, C. W.; Davie, D. C.; Charles, D. A.

    2015-12-01

    Geological disposal of nuclear waste is being increasingly considered to deal with the growing volume of waste resulting from the nuclear legacy of numerous nations. Within the UK there is 650,000 cubic meters of waste safely stored and managed in near-surface interim facilities but with no conclusive permanent disposal route. A Geological Disposal Facility with incorporated Engineered Barrier Systems are currently being considered as a permanent waste management solution (Fig.1). This research focuses on the EBS bentonite buffer/waste canister interface, and experimentally replicates key environmental phases that would occur after canister emplacement. This progresses understanding of the temporal evolution of the EBS and the associated impact on its engineering, mineralogical and physicochemical state and considers any consequences for the EBS safety functions of containment and isolation. Correlation of engineering properties to the physicochemical state is the focus of this research. Changes to geotechnical properties such as Atterberg limits, swelling pressure and swelling kinetics are measured after laboratory exposure to THMC variables from interface and batch experiments. Factors affecting the barrier, post closure, include corrosion product interaction, precipitation of silica, near-field chemical environment, groundwater salinity and temperature. Results show that increasing groundwater salinity has a direct impact on the buffer, reducing swelling capacity and plasticity index by up to 80%. Similarly, thermal loading reduces swelling capacity by 23% and plasticity index by 5%. Bentonite/steel interaction studies show corrosion precipitates diffusing into compacted bentonite up to 3mm from the interface over a 4 month exposure (increasing with temperature), with reduction in swelling capacity in the affected zone, probably due to the development of poorly crystalline iron oxides. These results indicate that groundwater conditions, temperature and corrosion

  4. Transuranic contaminated waste form characterization and data base

    International Nuclear Information System (INIS)

    Kniazewycz, B.G.; McArthur, W.C.

    1980-07-01

    This volume contains 5 appendices. Title listing are: technologies for recovery of transuranics; nondestructive assay of TRU contaminated wastes; miscellaneous waste characteristics; acceptance criteria for TRU waste; and TRU waste treatment technologies

  5. Biodegradation of uranium-contaminated waste oil

    International Nuclear Information System (INIS)

    Hary, L.F.

    1983-01-01

    The Portsmouth Gaseous Diffusion Plant routinely generates quantities of uranium-contaminated waste oil. The current generation rate of waste oil is approximately 2000 gallons per year. The waste is presently biodegraded by landfarming on open field soil plots. However, due to the environmental concerns associated with this treatment process, studies were conducted to determine the optimum biodegradation conditions required for the destruction of this waste. Tests using respirometric flasks were conducted to determine the biodegradation rate for various types of Portsmouth waste oil. These tests were performed at three different loading rates, and on unfertilized and fertilized soil. Additional studies were conducted to evaluate the effectiveness of open field landfarming versus treatment at a greenhouse-like enclosure for the purpose of maintaining soil temperatures above ambient conditions. The respirometric tests concluded that the optimum waste oil loading rate is 10% weight of oil-carbon/weight of soil (30,600 gallons of uranium-contaminated waste oil/acre) on soils with adjusted carbon:nitrogen and carbon:phosphorus ratios of 60:1 and 800:1, respectively. Also, calculational results indicated that greenhouse technology does not provide a significant increase in biodegradation efficiency. Based on these study results, a 6300 ft. 2 abandoned anaerobic digester sludge drying bed is being modified into a permanent waste oil biodegradation facility. The advantage of using this area is that uranium contamination will be contained by the bed's existing leachate collection system. This modified facility will be capable of handling approximately 4500 gallons of waste oil per year; accordingly current waste generation quantities will be satisfactorily treated. 15 refs., 14 figs., 4 tabs

  6. The prospective usage of the multi-purpose canister and impacts on the waste management and disposal system

    International Nuclear Information System (INIS)

    McLeod, N.B.

    1993-01-01

    The Multi-Purpose Canister (MPC) is designed to be loaded with spent fuel and sealed at reactors and then serve the functions of transport, storage and disposal without reopening. It can be either self-shielded or unshielded, thus requiring compatible overpacks for transport, storage and disposal. The MPC is not a new concept but it may now be viable because of the particular characteristics at Yucca Mountain: larger MPCs are possible because of ramp access to the repository horizon, and the less difficult temperature limits because of in-drift emplacement, rather than borehole emplacement. This paper describes the advantages and disadvantages of adopting the MPC as the principal technology to be employed in the US program. Use of the MPC permits integration of the utility and DOE portions of the system as well as among the elements within the DOE portion. Paradoxically, the principal disadvantage of the MPC is a direct consequence of its merit as an integrating technology. Full integration includes disposability without reopening, and requires that disposability design decisions be made and implemented well in advance of when waste package licensing uncertainties are resolved. There is, therefore, a risk that MPCs loaded prior to waste package licensing will have to be opened. This risk is discussed in terms of probability and consequences and various alternatives for mitigating this risk are discussed

  7. Treatment of contaminated waste plastics material

    International Nuclear Information System (INIS)

    Sims, J.; Hitchcock, J.W.

    1984-01-01

    Radioactive contaminated plastics material is treated by reducing it to uniform-sized debris and extruding it from a heated extruder into a sealed container in monolithic block form or as an in-fill matrix for other contaminated waste articles to create a substantially void-free sealed mass for disposal. Density adjusting fillers may be included. Extrusion may alternatively take place into a clean sealable plastics tube. (author)

  8. Management of Hazardous Waste and Contaminated Land

    OpenAIRE

    Hilary Sigman; Sarah Stafford

    2010-01-01

    Regulation of hazardous waste and cleanup of contaminated sites are two major components of modern public policy for environmental protection. We review the literature on these related areas, with emphasis on empirical analyses. Researchers have identified many behavioral responses to regulation of hazardous waste, including changes in the location of economic activity. However, the drivers behind compliance with these costly regulations remain a puzzle, as most research suggests a limited ro...

  9. Bioremediation of uranium contaminated soils and wastes

    International Nuclear Information System (INIS)

    Francis, A.J.

    1998-01-01

    Contamination of soils, water, and sediments by radionuclides and toxic metals from uranium mill tailings, nuclear fuel manufacturing and nuclear weapons production is a major concern. Studies of the mechanisms of biotransformation of uranium and toxic metals under various microbial process conditions has resulted in the development of two treatment processes: (1) stabilization of uranium and toxic metals with reduction in waste volume and (2) removal and recovery of uranium and toxic metals from wastes and contaminated soils. Stabilization of uranium and toxic metals in wastes is accomplished by exploiting the unique metabolic capabilities of the anaerobic bacterium, Clostridium sp. The radionuclides and toxic metals are solubilized by the bacteria directly by enzymatic reductive dissolution, or indirectly due to the production of organic acid metabolites. The radionuclides and toxic metals released into solution are immobilized by enzymatic reductive precipitation, biosorption and redistribution with stable mineral phases in the waste. Non-hazardous bulk components of the waste volume. In the second process uranium and toxic metals are removed from wastes or contaminated soils by extracting with the complexing agent citric acid. The citric-acid extract is subjected to biodegradation to recover the toxic metals, followed by photochemical degradation of the uranium citrate complex which is recalcitrant to biodegradation. The toxic metals and uranium are recovered in separate fractions for recycling or for disposal. The use of combined chemical and microbiological treatment process is more efficient than present methods and should result in considerable savings in clean-up and disposal costs

  10. Evaluation of a molybdenum assay canister

    International Nuclear Information System (INIS)

    Yoshizumi, T.T.; Keener, S.J.

    1988-01-01

    The performance characteristics of a commercial molybdenum assay canister were evaluated. The geometrical variation of the technetium-99m (/sup 99m/Tc) activity reading was studied as a function of the elution volume for the standard vials. It was found that the /sup 99m/Tc canister activity reading was ∼ 5% lower than that of the standard method. This is due to attenuation by the canister wall. However, the effect of the geometric variation on the clinical dose preparation was found to be insignificant. The molybdenum-99 ( 99 Mo) contamination level was compared by two methods: (1) the commercial canister and (2) the standard assay kit. The 99 Mo contamination measurements with the canister indicated consistently lower readings than those with the standard 99 Mo assay kit. The authors conclude that the canister may be used in the clinical settings. However, the user must be aware of the problems and the limitations associated with this canister

  11. Cesium release from ceramic waste form materials in simulated canister corrosion product containing solutions

    Energy Technology Data Exchange (ETDEWEB)

    Vittorio, Luca; Drabarek, Elizabeth; Chronis, Harriet; Griffith, Christopher S

    2004-07-01

    It has previously been demonstrated that immobilization of Cs{sup +} and/or Sr{sup 2+} sorbed on hexagonal tungsten oxide bronze (HTB) adsorbent materials can be achieved by heating the materials in air at temperatures in the range 500 - 1300 deg C. Highly crystalline powdered HTB materials formed by heating at 800 deg C show leach characteristics comparable to Cs-containing hot-pressed hollandites in the pH range from 0 to 12. As a very harsh leaching test, and also to model in a basic manner, leaching in the presence of canister corrosion products in oxidising environments, leaching of the bronzoid phases has been undertaken in Fe(NO{sub 3}){sub 3} solutions of increasing concentration. This is done in comparison with Cs -hollandite materials in order to compare the leaching characteristics of these two materials under such conditions. Both the Cs-loaded bronze and hollandite materials leach severely in Fe(NO{sub 3}){sub 3} losing virtually all of the immobilized Cs in a period of four days at 150 deg C. Total release of Cs and conversion of hollandite to titanium and iron titanium oxides begins to be observed at relatively low concentrations and is virtually complete after four days reaction in 0.5 mol/L Fe(NO{sub 3}){sub 3}. In the case of the bronze, all of the Cs is also extracted but the HTB structure is preserved. The reaction presumably involves an ion-exchange mechanism and iron oxide with a spinel structure is also observed at high Fe concentrations. (authors)

  12. Cesium release from ceramic waste form materials in simulated canister corrosion product containing solutions

    International Nuclear Information System (INIS)

    Vittorio, Luca; Drabarek, Elizabeth; Chronis, Harriet; Griffith, Christopher S.

    2004-01-01

    It has previously been demonstrated that immobilization of Cs + and/or Sr 2+ sorbed on hexagonal tungsten oxide bronze (HTB) adsorbent materials can be achieved by heating the materials in air at temperatures in the range 500 - 1300 deg C. Highly crystalline powdered HTB materials formed by heating at 800 deg C show leach characteristics comparable to Cs-containing hot-pressed hollandites in the pH range from 0 to 12. As a very harsh leaching test, and also to model in a basic manner, leaching in the presence of canister corrosion products in oxidising environments, leaching of the bronzoid phases has been undertaken in Fe(NO 3 ) 3 solutions of increasing concentration. This is done in comparison with Cs -hollandite materials in order to compare the leaching characteristics of these two materials under such conditions. Both the Cs-loaded bronze and hollandite materials leach severely in Fe(NO 3 ) 3 losing virtually all of the immobilized Cs in a period of four days at 150 deg C. Total release of Cs and conversion of hollandite to titanium and iron titanium oxides begins to be observed at relatively low concentrations and is virtually complete after four days reaction in 0.5 mol/L Fe(NO 3 ) 3 . In the case of the bronze, all of the Cs is also extracted but the HTB structure is preserved. The reaction presumably involves an ion-exchange mechanism and iron oxide with a spinel structure is also observed at high Fe concentrations. (authors)

  13. Technical note 2. A review of the creep ductility of copper for nuclear waste canister application

    International Nuclear Information System (INIS)

    Pettersson, Kjell

    2011-03-01

    Background: The Swedish Radiation Safety Authority (SSM) reviews the Swedish Nuclear Fuel Company's (SKB) applications under the Act on Nuclear Activities (SFS 1984:3) for the construction and operation of a repository for spent nuclear fuel and for an encapsulation facility. As part of the review, SSM commissions consultants to carry out work in order to obtain information on specific issues. The results from the consultants' tasks are reported in SSM's Technical Note series. Objectives of the project: This project is part of SSM:s review of SKB:s license application for final disposal of spent nuclear fuel. The assignment concerns review of creep mechanisms for copper material used as a corrosion barrier in canisters for nal disposal of nuclear fuel in Sweden. Summary by the author: SKB has presented insufficient evidence to justify their position that the OFP copper has an adequate creep ductility during long term storage. Their large body of experiments only serves to prove that the creep ductility is sufficient for much shorter time spans than the intended storage times. There is a clear need for a credible theory of creep brittleness of OFP copper which will permit extrapolations to long term storage. The theory presented by SKB does not in its present state permit credible extrapolations. Alternatively SKB needs to find an explanation to the effect of phosphorus on the creep ductility and that it ensures the absence of creep brittleness in OFP copper. It is interesting to note that SKB has presented experimental evidence that intergranular cracks can form in OFP material tested in cracked specimens. Perhaps it is possible to more systematically study formation and growth of intergranular cracks in specimens of OFP copper with cracks

  14. Press to compress contaminated wastes drums

    International Nuclear Information System (INIS)

    Prevost, J.

    1993-01-01

    This patent describes a press for contaminated wastes drums pressing. The press is made of a structure comprising a base and an upper stringer bind to the base by vertical bearers, a compression system comprising a main cylinder and a ram, connected to the upper stringer

  15. Contaminant transport at a waste residue deposit

    DEFF Research Database (Denmark)

    Engesgaard, Peter Knudegaard; Traberg, Rikke

    1996-01-01

    Contaminant transport in an aquifer at an incinerator waste residue deposit in Denmark is simulated. A two-dimensional, geochemical transport code is developed for this purpose and tested by comparison to results from another code, The code is applied to a column experiment and to the field site...

  16. Native copper in Permian Mudstones from South Devon: A natural analogue of copper canisters for high-level radioactive waste

    International Nuclear Information System (INIS)

    Milodowski, A.E.; Styles, M.T.; Werme, L.; Oversby, V.M.

    2001-01-01

    Native copper (>99.9% Cu) sheets associated with complex uraniferous and vanadiferous concretions in Upper Permian Mudstones from south Devon (United Kingdom) have been studied as a 'natural analogue' for copper canisters designed to be used in the isolation of spent fuel and high-level radioactive wastes (HLW) for deep geological disposal. Detailed analysis demonstrates that the copper formed before the mudstones were compacted. The copper displays complex corrosion and alteration. The earliest alteration was to copper oxides, followed sequentially by the formation of copper arsenides, nickel arsenide and copper sulphide, and finally nickel arsenide accompanied by nickel-copper arsenide, copper arsenide and uranium silicates. Petrographic observations demonstrate that these alteration products also formed prior to compaction. Consideration of the published history for the region indicates that maximum compaction of the rocks will have occurred by at least the Lower Jurassic (i.e. over 176 Ma ago). Since that time the copper sheets have remained isolated by the compacted mudstones and were unaffected by further corrosion until uplift and exposure to present-day surface weathering

  17. 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

  18. Hydrothermal processing of transuranic contaminated combustible waste

    International Nuclear Information System (INIS)

    Buelow, S.J.; Worl, L.; Harradine, D.; Padilla, D.; McInroy, R.

    2001-01-01

    Experiments at Los Alamos National Laboratory have demonstrated the usefulness of hydrothermal processing for the disposal of a wide variety of transuranic contaminated combustible wastes. This paper provides an overview of the implementation and performance of hydrothermal treatment for concentrated salt solutions, explosives, propellants, organic solvents, halogenated solvents, and laboratory trash, such as paper and plastics. Reaction conditions vary from near ambient temperatures and pressure to over 1000degC and 100 MPa pressure. Studies involving both radioactive and non-radioactive waste simulants are discussed. (author)

  19. Radiological hazards of alpha-contaminated waste

    International Nuclear Information System (INIS)

    Rodgers, J.C.

    1982-01-01

    The radiological hazards of alpha-contaminated wastes are discussed in this overview in terms of two components of hazard: radiobiological hazard, and radioecological hazard. Radiobiological hazard refers to human uptake of alpha-emitters by inhalation and ingestion, and the resultant dose to critical organs of the body. Radioecological hazard refers to the processes of release from buried wastes, transport in the environment, and translocation to man through the food chain. Besides detailing the sources and magnitude of hazards, this brief review identifies the uncertainties in their estimation, and implications for the regulatory process

  20. 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

  1. 32-Week Holding-Time Study of SUMMA Polished Canisters and Triple Sorbent Traps Used To Sample Organic Constituents in Radioactive Waste Tank Vapor Headspace

    International Nuclear Information System (INIS)

    Evans, John C.; Huckaby, James L.; Mitroshkov, Alexandre V.; Julya, Janet L.; Hayes, James C.; Edwards, Jeffrey A.; Sasaki, Leela M.

    1997-01-01

    Two sampling methods[SUMMA polished canisters and triple sorbent traps (TSTs)] were compared for long-term storage of trace organic vapor samples collected from the headspaces of high-level radioactive waste tanks at the U.S. Department of Energy's Hanford Site in Washington State. Because safety, quality assurance, radiological controls, the long-term stability of the sampling media during storage needed to be addressed. Samples were analyzed with a gas chromatograph/mass spectrometer (GC/MS) using cryogenic reconcentration or thermal desorption sample introduction techniques. SUMMA canister samples were also analyzed for total non-methane organic compounds (TNMOC) by GC/flame ionization detector (FID) using EPA Compendium Method TO-12 . To verify the long-term stability of the sampling media, multiple samples were collected in parallel from a typical passively ventilated radioactive waste tank known to contain moderately high concentrations of both polar and nonpolar organic compounds. Analyses for organic analytes and TNMOC were conducted at increasing intervals over a 32-week period to determine whether any systematic degradation of sample integrity occurred. Analytes collected in the SUMMA polished canisters generally showed good stability over the full 32 weeks with recoveries at the 80% level or better for all compounds studied. The TST data showed some loss (50-80% recovery) for a few high-volatility compounds even in the refrigerated samples; losses for unrefrigerated samples were far more pronounced with recoveries as low as 20% observed in a few cases

  2. Canister disposition plan for the DWPF Startup Test Program

    International Nuclear Information System (INIS)

    Harbour, J.R.; Payne, C.H.

    1990-01-01

    This report details the disposition of canisters and the canistered waste forms produced during the DWPF Startup Test Program. The six melter campaigns (DWPF Startup Tests FA-13, WP-14, WP-15, WP-16, WP-17, and FA-18) will produce 126 canistered waste forms. In addition, up to 20 additional canistered waste forms may be produced from glass poured during the transition between campaigns. In particular, this canister disposition plan (1) assigns (by alpha-numeric code) a specific canister to each location in the six campaign sequences, (2) describes the method of access for glass sampling on each canistered waste form, (3) describes the nature of the specific tests which will be carried out, (4) details which tests will be carried out on each canistered waste form, (5) provides the sequence of these tests for each canistered waste form, and (6) assigns a storage location for each canistered waste form. The tests are designed to provide evidence, as detailed in the Waste Form Compliance Plan (WCP 1 ), that the DWPF product will comply with the Waste Acceptance Product Specifications (WAPS 2 ). The WAPS must be met before the canistered waste form is accepted by DOE for ultimate disposal at the Federal Repository. The results of these tests will be included in the Waste Form Qualification Report (WQR)

  3. Alpha-contaminated waste management workshop

    International Nuclear Information System (INIS)

    1982-12-01

    These proceedings are published to provide a record of the oral presentations made at the DOE Alpha-Contaminated Workshop held in Gaithersburg, Maryland, on August 10-13, 1982. The papers are transcriptions of these oral presentations and, as such, do not contain as significant detail as will be found in the reviewed papers to be published in the periodical Nuclear and Chemical Waste Management in the first issue for 1983. These transcriptions have been reviewed by the speakers and some illustrations have been provided, but these contain only the preliminary information that will be provided in the technical papers to be published in the periodical. These papers have been grouped under the following headings: source terms; disposal technology and practices for alpha-contaminated waste; risk analyses and safety assessments. These papers in addition to those dealing with legislative and regulatory aspects have been abstracted and indexed for the Energy Data Base

  4. Conceptual designs of radioactive canister transporters

    International Nuclear Information System (INIS)

    1978-02-01

    This report covers conceptual designs of transporters for the vertical, horizontal, and inclined installation of canisters containing spent-fuel elements, high-level waste, cladding waste, and intermediate-level waste (low-level waste is not discussed). Included in the discussion are cask concepts; transporter vehicle designs; concepts for mechanisms for handling and manipulating casks, canisters, and concrete plugs; transporter and repository operating cycles; shielding calculations; operator radiation dosages; radiation-resistant materials; and criteria for future design efforts

  5. Incineration process for plutonium-contaminated waste

    International Nuclear Information System (INIS)

    Vincent, J.J.; Longuet, T.; Cartier, R.; Chaudon, L.

    1992-01-01

    A reprocessing plant with an annual throughput of 1600 metric tons of fuel generates 50 m 3 of incinerable α-contaminated waste. The reference treatment currently adopted for these wastes is to embed them in cement grout, with a resulting conditioned waste volume of 260 m 3 . The expense of mandatory geological disposal of such volumes justifies examination of less costly alternative solutions. After several years of laboratory and inactive pilot-scale research and development, the Commissariat a l'Energie Atomique has developed a two-step incineration process that is particularly suitable for α-contaminated chlorinated plastic waste. A 4 kg-h -1 pilot unit installed at the Marcoule Nuclear Center has now logged over 3500 hours in operation, during which the operating parameters have been optimized and process performance characteristics have been determined. Laboratory research during the same period has also determined the volatility of transuranic nuclides (U, Am and Pu) under simulated incineration conditions. A 100 g-h -1 laboratory prototype has been set up to obtain data for designing the industrial pilot facility

  6. Management of alpha-contaminated wastes

    International Nuclear Information System (INIS)

    1980-01-01

    Full text: With the increasing use of nuclear energy throughout the world for the generation of electric power and, especially, the use of the plutonium cycle for fast breeder reactors (FBRs), close attention has to be given to the safe management of alpha-contaminated wastes arising from spent-fuel reprocessing or mixed fuel fabrication Appropriate handling, conditioning and disposal of these wastes is, therefore, an activity of highest importance to ensure adequate protection of man and his environment from the potential hazard they pose over long periods of time. As the generation of alpha-contaminated waste is expected to increase considerably in the 1990s, when FBRs and the associated plutonium recycling will reach an industrial scale, it was felt timely to review the present state of the art in this area. The symposium organized jointly by the IAEA and the Commission of European Communities (CEC) was the first international symposium dealing with this specific topic. Its principle aim was to serve as 'zero-point' stating the present technical knowledge in view of the future needs for the management of alpha-contaminated wastes, before an industrial scale of production will be reached. The programme of the symposium was drawn up in eight sessions and covered the following topics: general policies; general practices; volume reduction techniques (two sessions); conditioning; alpha-monitoring; actinides partitioning; and disposal options. A variety of techniques has been investigated in various countries for several years for managing alpha-contaminated wastes. The first target was to reduce the volume of the wastes and to study matrices for the immobilization of waste radionuclides with a view to final waste disposal. At present, operational experience has been gained at different nuclear laboratories and facilities. At the same time various disposal options have been investigated. Some of the major items discussed at the symposium might be concluded as follows

  7. Soil contamination adjacent to waste tank 8

    International Nuclear Information System (INIS)

    Odum, J.V.

    1976-11-01

    In March and April 1961, miscalibrated liquid level instrumentation resulted in an overfilling of tank 8 to about 5 in. above the fill-line entrance. The resultant liquid head caused waste to seep through an asbestos-packed sleeve to the fill-line encasement and from there into the main encasement. Most of this waste returned to primary containment (i.e., the catch tank) through a separately encased drain line. However, approximately 1500 gal of high heat waste leaked from the fill-line encasement into the ground, probably through the joint at the juncture of the fill-line encasement and the concrete encasement of the waste tank. The contamination is contained in a 1000- to 1500-ft 3 zone of soil 12 to 26 ft below grade, 18 ft above the maximum elevation of the water table, and distributed roughly symmetrically around the fill-line encasement. Estimates from a continuing monitoring program indicate that less than 5000 Ci of 137 Cs, less than 0.005 Ci of 238 239 Pu, and less than 0.5 Ci of 89 90 Sr are in the soil. Analysis indicates that the contamination presents no current or future hazard to the environment; consequently, there is no technical reason for excavation of this soil. The high cost of excavation and exposure of personnel make excavation undesirable. The contaminated soil will remain under surveillance and undisturbed at tank 8 until the tank is removed from service, at which time its disposition will be re-evaluated

  8. DOE's plan for buried transuranic (TRU) contaminated waste

    International Nuclear Information System (INIS)

    Mathur, J.; D'Ambrosia, J.; Sease, J.

    1987-01-01

    Prior to 1970, TRU-contaminated waste was buried as low-level radioactive waste. In the Defense Waste Management Plan issued in 1983, the plan for this buried TRU-contaminated waste was to monitor the buried waste, take remedial actions, and to periodically evaluate the safety of the waste. In March 1986, the General Accounting Office (GAO) recommended that the Department of Energy (DOE) provide specific plans and cost estimates related to buried TRU-contaminated waste. This plan is in direct response to the GAO request. Buried TRU-contaminated waste and TRU-contaminated soil are located in numerous inactive disposal units at five DOE sites. The total volume of this material is estimated to be about 300,000 to 500,000 m 3 . The DOE plan for TRU-contaminated buried waste and TRU-contaminated soil is to characterize the disposal units; assess the potential impacts from the waste on workers, the surrounding population, and the environment; evaluate the need for remedial actions; assess the remedial action alternatives; and implement and verify the remedial actions as appropriate. Cost estimates for remedial actions for the buried TRU-contaminated waste are highly uncertain, but they range from several hundred million to the order of $10 billion

  9. Galvanic corrosion of copper-cast iron couples in relation to the Swedish radioactive waste canister concept

    International Nuclear Information System (INIS)

    Smart, N.R.; Fennell, P.A.H.; Rance, A.P.; Werme, L.O.

    2004-01-01

    To ensure the safe encapsulation of spent nuclear fuel rods for geological disposal, SKB are considering using the Copper-Iron Canister, which consists of an outer copper canister and an inner cast iron container. The canister will be placed into boreholes in the bedrock of a geologic repository and surrounded by bentonite clay. In the unlikely event of the outer copper canister being breached, water could enter the annulus between the inner and outer canister and at points of contact between the two metals there would be a possibility of galvanic interactions. To study this effect, copper-cast iron galvanic couples were set up in a number of different environments representing possible conditions in the SKB repository. The tests investigated two artificial pore-waters and a bentonite slurry, under aerated and deaerated conditions, at 30 deg. C and 50 deg. C. The currents passing between the coupled electrodes and the potential of the couples were monitored for several months. In addition, some bimetallic crevice specimens based on the multi-crevice assembly (MCA) design were used to simulate the situation where the copper canister will be in direct contact with the cast iron inner vessel. The effect of growing an oxide film on the surface of the cast iron prior to coupling it with copper was also investigated. The electrochemical results are presented graphically in the form of electrode potentials and galvanic corrosion currents as a function of time. The galvanic currents in aerated conditions were much higher than in deaerated conditions. For example, at 30 deg. C, galvanic corrosion rates as low as 0.02 μm/year were observed for iron in groundwater after de-aeration, but of the order of 100 μm/year for the cast iron at 50 deg. C in the presence of oxygen. The galvanic currents were generally higher at 50 deg. C than at 30 deg. C. None of the MCA specimens exhibited any signs of crevice corrosion under deaerated conditions. It will be shown that in deaerated

  10. Hydrothermal processing of actinide contaminated organic wastes

    International Nuclear Information System (INIS)

    Worl, A.; Buelow, S.J.; Le, L.A.; Padilla, D.D.; Roberts, J.H.

    1997-01-01

    Hydrothermal oxidation is an innovative process for the destruction of organic wastes, that occurs above the critical temperature and pressure of water. The process provides high destruction and removal efficiencies for a wide variety of organic and hazardous substances. For aqueous/organic mixtures, organic materials, and pure organic liquids hydrothermal processing removes most of the organic and nitrate components (>99.999%) and facilitates the collection and separation of the actinides. We have designed, built and tested a hydrothermal processing unit for the removal of the organic and hazardous substances from actinide contaminated liquids and solids. Here we present results for the organic generated at the Los Alamos National Laboratory Plutonium Facility

  11. Treatment of animal wastes contaminated with radioisotopes

    International Nuclear Information System (INIS)

    Morikawa, Naotake

    1979-01-01

    With increase of isotope utilizations as tracers in medicine, pharmacy, agriculture, biology and others, the management of resultant organic waste liquids and animal wastes is becoming a major problem. For the animal wastes contaminated with radioisotopes, numbers of studies and tests showed that drying them fully and the subsequent suitable disposal would be the most feasible procedures. This new method is being carried out since last year, which will shortly take the place of the keeping in formalin. For the drying, two alternative processes in particular are being investigated. As the one, freeze-drying apparatuses consist of refrigerating and freeze-drying devices. As the other, microwave-drying apparatuses feature rapid dehydration. The following matters are described: problems emerged in the course of studies and test; the drying processes, i.e. freeze-drying and microwave-drying, and their respective characteristics; and views of the Nuclear Safety Bureau, Science and Technology Agency, on animal waste drying. (J.P.N.)

  12. Transuranic contaminated waste functional definition and implementation

    International Nuclear Information System (INIS)

    Kniazewycz, B.G.

    1980-03-01

    The purpose of this report is to examine the problem(s) of TRU waste classification and to document the development of an easy-to-apply standard(s) to determine whether or not this waste package should be emplaced in a geologic repository for final disposition. Transuranic wastes are especially significant because they have long half-lives and some are rather radiotoxic. Transuranic radionuclides are primarily produced by single or multiple neutron capture by U-238 in fuel elements during the operation of a nuclear reactor. Reprocessing of spent fuel elements attempts to remove plutonium, but since the separation is not complete, the resulting high-activity liquids still contain some plutonium as well as other transuranics. Likewise, transuranic contamination of low-activity wastes also occurs when the transuranic materials are handled or processed, which is primarily at federal facilities involved in R and D and nuclear weapons production. Transuranics are persistent in the environment and, as a general rule, are strongly retained by soils. They are not easily transported through most food chains, although some reconcentration does take place in the aquatic food chain. They pose no special biological hazard to humans upon ingestion because they are weakly absorbed from the gastrointestional tract. A greater hazard results from inhalation since they behave like normal dust and fractionate accordingly

  13. Defense Waste Management Plan for buried transuranic-contaminated waste, transuranic-contaminated soil, and difficult-to-certify transuranic waste

    International Nuclear Information System (INIS)

    1987-06-01

    GAO recommended that DOE provide specific plans for permanent disposal of buried TRU-contaminated waste, TRU-contaminated soil, and difficult-to-certify TRU waste; cost estimates for permanent disposal of all TRU waste, including the options for the buried TRU-contaminated waste, TRU-contaminated soil, and difficult-to-certify TRU waste; and specific discussions of environmental and safety issues for the permanent disposal of TRU waste. Purpose of this document is to respond to the GAO recommendations by providing plans and cost estimates for the long-term isolation of the buried TRU-contaminated waste, TRU-contaminated soil, and difficult-to-certify TRU waste. This report also provides cost estimates for processing and certifying stored and newly generated TRU waste, decontaminating and decommissioning TRU waste processing facilities, and interim operations

  14. Recovery of Mercury From Contaminated Liquid Wastes

    International Nuclear Information System (INIS)

    1998-01-01

    The Base Contract program emphasized the manufacture and testing of superior sorbents for mercury removal, testing of the sorption process at a DOE site, and determination of the regeneration conditions in the laboratory. During this project, ADA Technologies, Inc. demonstrated the following key elements of a successful regenerable mercury sorption process: (1) sorbents that have a high capacity for dissolved, ionic mercury; (2) removal of ionic mercury at greater than 99% efficiency; and (3) thermal regeneration of the spent sorbent. ADA's process is based on the highly efficient and selective sorption of mercury by noble metals. Contaminated liquid flows through two packed columns that contain microporous sorbent particles on which a noble metal has been finely dispersed. A third column is held in reserve. When the sorbent is loaded with mercury to the point of breakthrough at the outlet of the second column, the first column is taken off-line and the flow of contaminated liquid is switched to the second and third columns. The spent column is regenerated by heating. A small flow of purge gas carries the desorbed mercury to a capture unit where the liquid mercury is recovered. Laboratory-scale tests with mercuric chloride solutions demonstrated the sorbents' ability to remove mercury from contaminated wastewater. Isotherms on surrogate wastes from DOE's Y-12 Plant in Oak Ridge, Tennessee showed greater than 99.9% mercury removal. Laboratory- and pilot-scale tests on actual Y-12 Plant wastes were also successful. Mercury concentrations were reduced to less than 1 ppt from a starting concentration of 1,000 ppt. The treatment objective was 50 ppt. The sorption unit showed 10 ppt discharge after six months. Laboratory-scale tests demonstrated the feasibility of sorbent regeneration. Results show that sorption behavior is not affected after four cycles

  15. Effects of radiation on the chemical environment surrounding waste canisters in proposed repository sites and possible effects on the corrosion process

    International Nuclear Information System (INIS)

    Glass, R.S.

    1981-12-01

    This report explores the interaction of ionizing radiation with various environments. In particular, worst case (aqueous) environments for the proposed nuclear waste repository sites are considered. Emphasis is on the fundamental chemical and physical processes involved. The identities of possible radiolysis products (both transient and stable) have been sought through a literature search. The effect of radiation on corrosion processes is discussed. The radiation-induced chemical environment in the worst case repository sites is not well defined. Attention should therefore be given to fundamental studies exploring the interaction of such environments with components of the nuclear waste package, including the canister materials and backfills. Identification and quantification of radiolysis products would be helpful in this regard

  16. Incineration of simulated plutonium-contaminated waste

    International Nuclear Information System (INIS)

    Ford, L.H.; Jenkins, M.J.

    1984-01-01

    Pyrolysis rate data are presented which will enable larger pyrolyser furnaces to be made for processing solid plutonium-contaminated materials at throughputs of up to 20 kg/h using either 1 or 2.5 kg packages as feed. The influence of liquids, such as water, kerosene or oil, on the pyrolysis process has also been assessed. The products of pyrolysis for a range of individual materials and their mixtures have been defined. The oxidation rates for both static and stirred beds of char have been obtained. The implications of both the pyrolysis and char-oxidation processes for plant design are discussed. This work has been commissioned by the Department of the Environment as part of its radioactive waste management programme. The results will be used in the formulation of government policy, but as this stage they do not necessarily represent that policy

  17. Processing and discarding method for contaminated concrete wastes

    International Nuclear Information System (INIS)

    Yamamoto, Kazuo; Konishi, Masao; Matsuda, Atsuo; Iwamoto, Yoshiaki; Yoshikane, Toru; Koie, Toshio; Nakajima, Yoshiro

    1998-01-01

    Contaminated concrete wastes are crashed into granular concrete wastes having a successive grain size distribution. They are filled in a contamination processing vessel and made hardenable in the presence of a water-hardenable material in the granular concrete wastes. When underground water intrudes into the contamination processing vessel filled with the granular concrete wastes upon long-term storage, the underground water reacts with the water-hardenable material to be used for the solidification effect. Accordingly, leaching of contaminated materials due to intrusion of underground water can be suppressed. Since the concrete wastes have a successive grain size distribution, coarse grains can be used as coarse aggregates, medium grains can be used as fine aggregates and fine grains can be used as a solidifying material. Accordingly, the amount of wastes after processing can be remarkably reduced, with no supply of a solidifying material from outside. (T.M.)

  18. Mass transfer between waste canister and water seeping in rock fractures. Revisiting the Q-equivalent model

    International Nuclear Information System (INIS)

    Neretnieks, Ivars; Liu Longcheng; Moreno, Luis

    2010-03-01

    Models are presented for solute transport between seeping water in fractured rock and a copper canister embedded in a clay buffer. The migration through an undamaged buffer is by molecular diffusion only as the clay has so low hydraulic conductivity that water flow can be neglected. In the fractures and in any damaged zone seeping water carries the solutes to or from the vicinity of the buffer in the deposition hole. During the time the water passes the deposition hole molecular diffusion aids in the mass transfer of solutes between the water/buffer interface and the water at some distance from the interface. The residence time of the water and the contact area between the water and the buffer determine the rate of mass transfer between water and buffer. Simple analytical solutions are presented for the mass transfer in the seeping water. For complex migration geometries simplifying assumptions are made that allow analytical solutions to be obtained. The influence of variable apertures on the mass transfer is discussed and is shown to be moderate. The impact of damage to the rock around the deposition hole by spalling and by the presence of a cemented and fractured buffer is also explored. These phenomena lead to an increase of mass transfer between water and buffer. The overall rate of mass transfer between the bulk of the water and the canister is proportional to the overall concentration difference and inversely proportional to the sum of the mass transfer resistances. For visualization purposes the concept of equivalent flowrate is introduced. This entity can be thought as of the flowrate of water that will be depleted of its solute during the water passage past the deposition hole. The equivalent flowrate is also used to assess the release rate of radionuclides from a damaged canister. Examples are presented to illustrate how various factors influence the rate of mass transfer

  19. Radioactive waste management for a radiologically contaminated hospitalized patient

    International Nuclear Information System (INIS)

    Pina Jomir, G.; Michel, X.; Lecompte, Y.; Chianea, N.; Cazoulat, A.

    2015-01-01

    Radioactive waste management in the post-accidental phase following caring for a radiologically contaminated patient in a hospital decontamination facility must be anticipated at a local level to be truly efficient, as the volume of waste could be substantial. This management must comply with the principles set out for radioactive as well as medical waste. The first step involves identification of radiologically contaminated waste based on radioactivity measurement for volume reduction. Then, the management depends on the longest radioactive half-life of contaminative radionuclides. For a half-life inferior to 100 days, wastes are stored for their radioactivity to decay for at least 10 periods before disposal like conventional medical waste. Long-lived radioactive waste management implies treatment of liquid waste and special handling for sorting and packaging before final elimination at the French National Agency for Radioactive Waste Management (ANDRA). Following this, highly specialized waste management skills, financial responsibility issues and detention of non-medical radioactive sources are questions raised by hospital radioactive waste management in the post-accidental phase. (authors)

  20. Bonding of radioactive contamination. IV. Effect of surface finish

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1983-01-01

    The mechanisms by which radioactive contamination would be bonded to a DWPF canister are being investigated. Previous investigations in this series have examined the effects of temperature, oxidation before contamination, and atmosphere composition control on the bonding of contamination. This memorandum describes the results of tests to determine the effect of special surface finishes on the bonding of contamination to waste glass canisters. Surface pretreatments which produce smoother canister surfaces actually cause radioactive contamination to be more tightly bonded to the metal surface than on an untreated surface. Based on the results of these tests it is recommended that the canister surface finish be specified as having a bright cold rolled mill finish equivalent to ASTM No. 2B. 7 references, 3 figures, 3 tables

  1. Containment of transuranic contamination at the early waste retrieval project

    International Nuclear Information System (INIS)

    Harness, J.L.; McKinney, J.D.

    1977-01-01

    On July 26, 1976, while retrieving buried transuranic waste under the Early Waste Retrieval Program, a corroded 55-gallon 17H drum was retrieved. When uprighted, several liters of liquid escaped from the drum. This liquid was contaminated with transuranics, principally Pu-239, Am-241, and some Pu-238. As a result of the spread of this contamination in the Operating Area Confinement, six working days were required to decontaminate the area. At no time did the contamination escape the interior of the Operating Area Confinement building, and no contamination to personnel resulted from this occurrence, nor was a hazard presented to the general public. The facility was designed and constructed to contain the transuranic contamination resulting from such an occurrence. Proper prior planning and personnel training prevented the contamination occurrence from becoming a major event. This report details the occurrence, the recovery, and the information obtained from this event

  2. Process for treating waste water having low concentrations of metallic contaminants

    Science.gov (United States)

    Looney, Brian B; Millings, Margaret R; Nichols, Ralph L; Payne, William L

    2014-12-16

    A process for treating waste water having a low level of metallic contaminants by reducing the toxicity level of metallic contaminants to an acceptable level and subsequently discharging the treated waste water into the environment without removing the treated contaminants.

  3. Process for reducing radioactive contamination in waste product gypsum

    International Nuclear Information System (INIS)

    Lange, P.H. Jr.

    1979-01-01

    A process is described for reducing the radioactive contamination in waste product gypsum in which waste product gypsum is reacted with a dilute sulfuric acid containing barium sulfate to form an acid slurry at an elevated temperature, the slurry is preferably cooled, the acid component is separated from the solid, and the resulting solid is separated into a fine fraction and a coarse fraction. The fine fraction predominates in barium sulfate and radioactive contamination. The coarse fraction predominates in a purified gypsum product of reduced radioactive contamination

  4. 78 FR 46447 - Conditional Exclusions From Solid Waste and Hazardous Waste for Solvent-Contaminated Wipes

    Science.gov (United States)

    2013-07-31

    ... section 307 of the Clean Water Act (CWA)); A municipal solid waste landfill that is regulated under 40 CFR... laundries and dry cleaners could dispose of sludge from cleaning solvent-contaminated wipes in solid waste landfills if the sludge does not exhibit a hazardous waste characteristic. \\8\\ The Agency stated in the...

  5. Vitrification of transuranic and beta-gamma contaminated solid wastes

    International Nuclear Information System (INIS)

    Dukes, M.D.

    1980-06-01

    Vitrification of solid transuranic contaminated (TRU) wastes alone and with high-level liquid wastes (HLLW) was studied. Homogeneous glasses containing 20 to 30 wt % ash were made by using glass frits previously developed at the Savannah River Plant and Pacific Northwest Laboratories. If the ash is vitrified along with the HLLW, 1.0 wt % as can be added to the waste forms without affecting their quality. This loading of ash is well above the loading required by the relative amounts of HLLW and TRU ash that will be processed at the Savannah River Plant. Vitrification of TRU-contaminated electropolishing sludges and high efficiency particular air filter materials along with HLLW would require an increase in the quantity of glass to be produced. However, if these TRU-contaminated solids were vitrified with the HLLW, the addition of low-level beta-gamma contaminated ash would require no further increase in glass production

  6. Transuranic contaminated waste container characterization and data base. Revision I

    International Nuclear Information System (INIS)

    Kniazewycz, B.G.

    1980-05-01

    The Nuclear Regulatory Commission (NRC) is developing regulations governing the management, handling and disposal of transuranium (TRU) radioisotope contaminated wastes as part of the NRC's overall waste management program. In the development of such regulations, numerous subtasks have been identified which require completion before meaningful regulations can be proposed, their impact evaluated and the regulations implemented. This report was prepared to assist in the development of the technical data base necessary to support rule-making actions dealing with TRU-contaminated wastes. An earlier report presented the waste sources, characteristics and inventory of both Department of Energy (DOE) generated and commercially generated TRU waste. In this report a wide variety of waste sources as well as a large TRU inventory were identified. The purpose of this report is to identify the different packaging systems used and proposed for TRU waste and to document their characteristics. This document then serves as part of the data base necessary to complete preparation and initiate implementation of TRU waste container and packaging standards and criteria suitable for inclusion in the present TRU waste management program. It is the purpose of this report to serve as a working document which will be used as appropriate in the TRU Waste Management Program. This report, and those following, will be compatible not only in format, but also in reference material and direction

  7. Contamination control aspects of attaching waste drums to the WIPP Waste Characterization Chamber

    International Nuclear Information System (INIS)

    Rubick, L.M.; Burke, L.L.

    1998-01-01

    Argonne National Laboratory West (ANL-W) is verifying the characterization and repackaging of contact-handled transuranic (CH-TRU) mixed waste in support of the Waste Isolation Pilot Program (WIPP) project located in Carlsbad, New Mexico. The WIPP Waste Characterization Chamber (WCC) was designed to allow opening of transuranic waste drums for this process. The WCC became operational in March of 1994 and has characterized approximately 240 drums of transuranic waste. The waste drums are internally contaminated with high levels of transuranic radionuclides. Attaching and detaching drums to the glove box posed serious contamination control problems. Prior to characterizing waste, several drum attachment techniques and materials were evaluated. An inexpensive HEPA filter molded into the bagging material helps with venting during detachment. The current techniques and procedures used to attach and detach transuranic waste drums to the WCC are described

  8. General procedure to characterize hazardous waste contaminated with radionuclides

    International Nuclear Information System (INIS)

    Vokal, A.; Svoboda, K.; Necasova, M.

    2002-04-01

    The report is structured as follows: Overview of current status of characterization of hazardous wastes contaminated with radionuclides (HWCTR) in the Czech Republic (Legislative aspects; Categorization of HWCwR; Overview of HWCwR emerging from workplaces handling ionizing radiation sources; Mixed waste management in the Czech Republic); General procedure to characterized wastes of the HWCwR type (Information needed from the waste producer; Waste analysis plan - description of waste treatment facilities, verification of wastes, selection of waste parameters followed, selection of sampling method, selection of test methods, selection of frequency of analyses; Radiation protection plan; Non-destructive methods of verification of waste - radiography/tomography, dosimetric inspection, measuring instrumentation, methods usable for the determination of volume and surface activities of materials; Non-destructive invasive methods - internal pressure measurement and gas analysis, endoscopic examination, visual inspection; Destructive methods - sampling, current equipment at Nuclear Research Institute Rez; Identification of hazardous components in waste - chemical screening of mixed wastes; Assessment of immobilization waste matrices; Assessment of packaging; Safety analyses; QA and QC). (P.A.)

  9. Waste disposal package

    Science.gov (United States)

    Smith, M.J.

    1985-06-19

    This is a claim for a waste disposal package including an inner or primary canister for containing hazardous and/or radioactive wastes. The primary canister is encapsulated by an outer or secondary barrier formed of a porous ceramic material to control ingress of water to the canister and the release rate of wastes upon breach on the canister. 4 figs.

  10. Waste reduction by separation of contaminated soils during environmental restoration

    International Nuclear Information System (INIS)

    Roybal, J.A.; Conway, R.; Galloway, B.; Vinsant, E.; Slavin, P.; Guerin, D.

    1998-06-01

    During cleanup of contaminated sites, Sandia National Laboratories, New Mexico (SNL/NM) frequently encounters soils with low-level radioactive contamination. The contamination is not uniformly distributed, but occurs within areas of clean soil. Because it is difficult to characterize heterogeneously contaminated soils in detail and to excavate such soils precisely using heavy equipment, it is common for large quantities of uncontaminated soil to be removed during excavation of contaminated sites. This practice results in the commingling and disposal of clean and contaminated material as low-level waste (LLW), or possibly low-level mixed waste (LLMW). Until recently, volume reduction of radioactively contaminated soil depended on manual screening and analysis of samples, which is a costly and impractical approach and does not uphold As Low As Reasonably Achievable (ALARA) principles. To reduce the amount of LLW and LLMW generated during the excavation process, SNL/NM is evaluating two alternative technologies. The first of these, the Segmented Gate System (SGS), is an automated system that located and removes gamma-ray emitting radionuclides from a host matrix (soil, sand, dry sludge). The matrix materials is transported by a conveyor to an analyzer/separation system, which segregates the clean and contaminated material based on radionuclide activity level. The SGS was used to process radioactively contaminated soil from the excavation of the Radioactive Waste Landfill. The second technology, Large Area Gamma Spectroscopy (LAGS), utilizes a gamma spec analyzer suspended over a slab upon which soil is spread out to a uniform depth. A counting period of approximately 30 minutes is used to obtain a full-spectrum analysis for the isotopes of interest. The LAGS is being tested on the soil that is being excavated from the Classified Waste Landfill

  11. Development of thermal conditioning technology for alpha-contaminated wastes

    International Nuclear Information System (INIS)

    Kim, Joon Hyung; Kim, H. Y.; Kim, J. G.

    2001-04-01

    To develop a thermal conditioning technology for alpha-contaminated wastes, which are presumed to generate from pyrochemical processing of spent fuel, research on the three different fields have been performed; incineration, off-gas treatment, and vitrification/cementation technology. Through the assessment on the amount of alpha-contaminated waste and incineration characterises, an oxygen-enriched incineration process, which can greatly reduce the off-gas volume, was developed by our own technology. Trial burn test with paper waste resulted in a reduction of off-gas volume by 3.5. A study on the behavior and adsorption of nuclides/heavy metals at high-temperature was performed to develop an efficient removal technology. Off-gas treatment technologies for radioiodine at high-temperature and 14 CO 2 , acidic gases, and radioactive gaseous wastes such as Xe/Kr at room temperature were established. As a part of development of high-level waste solidification technology, manufacture of high-frequency induction melter, fabrication and characterization of base-glass media fabricated with spent HEPA filter medium, and development of titanate ceramic material as a precursor of SYNROC by a self-combustion method were performed. To develop alpha-contaminated waste solidification technology, a process to convert periodontal in the cement matrix to calcite with SuperCritical Carbon Dioxide (SCCD) was manufactured. The SCCD treatment enhanced the physicochemical properties of cement matrices, which increase the long-term integrity of cement waste forms during transportation and storage

  12. Gamma radiation sterilises bacteria-contaminated waste

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    The use of gamma radiation for the sterilisation of sewage and hospital waste etc., is briefly described. A sterilisation plant delivered by Sulzer is illustrated diagrammatically and its functioning explained, while a photograph illustrates a similar plant delivered by Geodel Systems, of Canada. The latter firm has adapted this type of plant for the radiolytical destruction of phenols, cyanides, alkyl benzenesulphonates and similar wastes. (JIW)

  13. Alpha-contaminated waste from reprocessing of nuclear fuel

    International Nuclear Information System (INIS)

    Sumner, W.

    1982-01-01

    The anticipated alpha-waste production rates from the Barnwell Nuclear Fuel Reprocessing plant is discussed. The estimated alpha-waste production rate from the 1500 metric ton/year plant is about 85,000 ft 3 /year at the 10 nCi/g limit. Most of this waste is estimated to come from the separation facility, and the major waste sources were cladding, which was 27%, and low-level contact-handled general process trash, which was estimated at 32% of the total. It was estimated that 45% of the waste was combustible and 72% of the waste was compactible. These characteristics could have a significant impact on the final volumes as disposed. Changing the alpha-waste limit from 10 nCi/g to 100 nCi/g was estimated to reduce the amount of alpha waste produced by about 20%. Again, the uncertainty in this value obviously has to be substantial. One has to recognize that these estimates were just that; they were not based on any operating experience. The total plutonium losses to waste, including the high-level waste, was estimated to be 1.5%. The cladding waste was estimated to be contaminated with alpha emitters to the extent of 10 4 to 10 5 nCi/g

  14. Long term management of wastes contaminated by plutonium

    International Nuclear Information System (INIS)

    Marque, Y.

    1983-01-01

    For the different categories of wastes, the evolution of the cumulated production until the year 2000 is described by curves and the general situation of production points is presented, all that in France. The storage conditions are specified according to the type of wastes, category A, B, or C; the threshold under which the waste is classified in A category being fixed by the safety authorities at 2.10 4 CMA (maximum permissible concentration), that is to say for plutonium 1Ci/m 3 . The knowledge of waste activity is another basic element of the management of such wastes, the fixing of the threshold, above which wastes contaminated by plutonium have to be stored underground, still keeping to be specified [fr

  15. Radiation safety for incineration of radioactive waste contaminated by cesium

    International Nuclear Information System (INIS)

    Veryuzhs'kij, Yu.V.; Gryin'ko, O.M.; Tokarevs'kij, V.V.

    2016-01-01

    Problems in the treatment of radioactive waste contaminated by cesium nuclides are considered in the paper. Chornobyl experience in the management of contaminated soil and contaminated forests is analyzed in relation to the accident at Fukushima-1. The minimization of release of cesium aerosols into atmosphere is very important. Radiation influence of inhaling atmosphere aerosols polluted by cesium has damage effect for humans. The research focuses on the treatment of forests contaminated by big volumes of cesium. One of the most important technologies is a pyro-gasification incineration with chemical reactions of cesium paying attention to gas purification problems. Requirements for process, physical and chemical properties of treatment of radioactive waste based on the dry pyro-gasification incineration facilities are considered in the paper together with the discussion of details related to incineration facilities. General similarities and discrepancies in the environmental pollution caused by the accidents at Chornobyl NPP and Fukushima-1 NPP in Japan are analyzed

  16. Tritium contaminated waste management at the tritium systems test assembly

    International Nuclear Information System (INIS)

    Jalbert, R.A.; Carlson, R.V.

    1987-01-01

    The Tritium Systems Test Assembly (TSTA) at Los Alamos continues to move toward full operation of an integrated, full-sized, computer-controlled fusion fuel processing loop. Concurrent nonloop experiments further the development of advanced tritium technologies and handling methods. Since tritium operations began in June 1984, tritium contaminated wastes have been produced at TSTA that are roughly typical in kind and amount of those to be produced by tritium fueling operations at fusion reactors. Methods of managing these wastes are described, including information on some methods of decontamination so that equipment can be reused. Data are given on the kinds and amounts of wastes and the general level of contamination. Also included are data on environmental emissions and doses to personnel that have resulted from TSTA operations. Particular problems in waste managements are discussed

  17. Distinguishing method for contamination/radio-activation of radioactive wastes

    International Nuclear Information System (INIS)

    Fukazawa, Takuji; Kato, Keiichiro; Koda, Satoshi.

    1994-01-01

    The present invention concerns a method of distinguishing the contamination/radio-activation of radioactive wastes used in processing wastes generated upon dismantling of exhausted nuclear reactors. Especially, contaminated/radio-activation is distinguished for wastes having openings such as pipes and valves, by utilizing scattering of γ-rays or γ-ray to β-ray ratio. That is, ratio of scattered γ-rays and direct γ-rays or ratio of β-rays and γ-rays from radioactive wastes are measured and compared by a radiation detector, to distinguish whether the radioactive wastes contaminated materials or radio-activated materials. For example, when an object to be measured having an opening is contaminated at the inner side, the radiation detector facing to the opening mainly detects high direct γ-rays emitted from the object to be measured while a radiation detector not facing the opening mainly detects high scattered γ-rays relatively. On the other hand, when the object is a radio-activated material, any of the detectors detect scattered γ-rays, so that they can be distinguished by these ratios. (I.S.)

  18. System to control contamination during retrieval of buried TRU waste

    Science.gov (United States)

    Menkhaus, Daniel E.; Loomis, Guy G.; Mullen, Carlan K.; Scott, Donald W.; Feldman, Edgar M.; Meyer, Leroy C.

    1993-01-01

    A system to control contamination during the retrieval of hazardous waste comprising an outer containment building, an inner containment building, within the outer containment building, an electrostatic radioactive particle recovery unit connected to and in communication with the inner and outer containment buildings, and a contaminate suppression system including a moisture control subsystem, and a rapid monitoring system having the ability to monitor conditions in the inner and outer containment buildings.

  19. Final Rule: 2013 Conditional Exclusions From Solid Waste and Hazardous Waste for Solvent-Contaminated Wipes

    Science.gov (United States)

    This is a regulation page for the final rule EPA issued on July 31, 2013 that modifies the hazardous waste management regulations for solvent-contaminated wipes under the Resource Conservation and Recovery Act (RCRA).

  20. Management of waste contaminated with alpha emitters

    International Nuclear Information System (INIS)

    Cartier, R.; Durec, J.P.

    1993-01-01

    Although advances are being made in the deep geological storage concept, it will probably never be possible to dispose of all types and quantities of radioactive waste in geological formations. Permanent storage should therefore only be considered as an option for final waste disposal. We are currently obliged to search for technological solutions which will reduce the quantities of waste to be managed by future generations, and to ensure that such management can be carried out safely without releasing elements detrimental to the environment into the biosphere. This clearly stated determination, combined with an attitude of complete openness, will secure public acceptance of nuclear energy. As a result of the Research and Development work described here, in March 1992 the Valduc Nuclear Research Center decided to build an industrial waste incineration facility. The facility was to have an annual incinerating capacity of 26 tonnes of waste with a mean radioactivity level of 7.5*10 8 Bq/kg (0.02 Ci/kg). Detailed design studies are in progress, procurements have been launched and construction of the building has started. Commercial operation is scheduled for late 1995. 4 refs. 2 figs

  1. Fluidized bed incineration of transuranic contaminated waste

    International Nuclear Information System (INIS)

    Ziegler, D.L.; Johnson, A.J.

    1978-01-01

    A 9 kg/hr pilot scale fluidized bed incinerator is now being used for burning various types of radioactive waste at Rocky Flats Plant. General solid combustible waste containing halogenated materials is burned in a fluidized bed of sodium carbonate for in situ neutralization of thermally generated acidic gases. A variety of other production related materials has been burned in the incinerator, including ion exchange resin, tributyl phosphate solutions, and air filters. Successful operation of the pilot plant incinerator has led to the design and construction of a production site unit to burn 82 kg/hr of plant generated waste. Residues from incinerator operations will be processed into glass buttons utilizing a vitrification plant now under development

  2. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    Hamilton, L.

    2001-01-01

    The Plutonium Immobilization Project (PIP) is a program funded by the U.S. Department of Energy to develop technology to disposition excess weapons grade plutonium. This program introduces the ''Can-in-Canister'' (CIC) technology that immobilizes the plutonium by encapsulating it in ceramic forms (or pucks) and ultimately surrounding it with high-level waste glass to provide a deterrent to recovery. Since there are significant radiation, contamination and security concerns, the project team is developing unique technologies to remotely perform plutonium immobilization tasks. This paper covers the design, development and testing of the magazines (cylinders containing cans of ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a jointed arm robot that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102 mm) diameter canister throat

  3. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    Hamilton, L.

    2001-01-01

    The Plutonium Immobilization Project (PIP) is a program funded by the U.S. Department of Energy to develop technology to disposition excess weapons grade plutonium. This program introduces the ''Can-in-Canister'' (CIC) technology that immobilizes the plutonium by encapsulating it in ceramic forms (or pucks) and ultimately surrounding it with high-level waste glass to provide a deterrent to recovery. Since there are significant radiation, contamination and security concerns, the project team is developing unique technologies to remotely perform plutonium immobilization tasks. This paper covers the design, development and testing of the magazines (cylinders containing cans of ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a join ted arm robot that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102 mm) diameter canister throat

  4. Toward zero waste events: Reducing contamination in waste streams with volunteer assistance.

    Science.gov (United States)

    Zelenika, Ivana; Moreau, Tara; Zhao, Jiaying

    2018-03-22

    Public festivals and events generate a tremendous amount of waste, especially when they involve food and drink. To reduce contamination across waste streams, we evaluated three types of interventions at a public event. In a randomized control trial, we examined the impact of volunteer staff assistance, bin tops, and sample 3D items with bin tops, on the amount of contamination and the weight of the organics, recyclable containers, paper, and garbage bins at a public event. The event was the annual Apple Festival held at the University of British Columbia, which was attended by around 10,000 visitors. We found that contamination was the lowest in the volunteer staff condition among all conditions. Specifically, volunteer staff reduced contamination by 96.1% on average in the organics bin, 96.9% in the recyclable containers bin, 97.0% in the paper bin, and 84.9% in the garbage bin. Our interventions did not influence the weight of the materials in the bins. This finding highlights the impact of volunteers on reducing contamination in waste streams at events, and provides suggestions and implications for waste management for event organizers to minimize contamination in all waste streams to achieve zero waste goals. Copyright © 2018. Published by Elsevier Ltd.

  5. Industrial-Scale Processes For Stabilizing Radioactively Contaminated Mercury Wastes

    International Nuclear Information System (INIS)

    Broderick, T. E.; Grondin, R.

    2003-01-01

    This paper describes two industrial-scaled processes now being used to treat two problematic mercury waste categories: elemental mercury contaminated with radionuclides and radioactive solid wastes containing greater than 260-ppm mercury. The stabilization processes were developed by ADA Technologies, Inc., an environmental control and process development company in Littleton, Colorado. Perma-Fix Environmental Services has licensed the liquid elemental mercury stabilization process to treat radioactive mercury from Los Alamos National Laboratory and other DOE sites. ADA and Perma-Fix also cooperated to apply the >260-ppm mercury treatment technology to a storm sewer sediment waste collected from the Y-12 complex in Oak Ridge, TN

  6. High temperature slagging incinerator for alpha contaminated wastes

    International Nuclear Information System (INIS)

    Van de Voorde, N.

    1985-01-01

    This report describes the experiences collected by the treatment of plutonium-contaminated wastes, in the High Temperature Slagging Incinerator at the C.E.N./S.C.K. at Mol, with the support of the Commission of the European Communities. The major objective of the exercise is to demonstrate the operability of this facility for the treatment of mixed transuranic (TRU) and beta-gamma solid waste material. The process will substantially reduce the TRU waste volume by burning the combustibles and converting the non-combustibles into a chemically inert and physically stable basalt-like slag product, suitable for safe transport and final disposal. (Auth.)

  7. Swedish recovered wood waste: linking regulation and contamination.

    Science.gov (United States)

    Krook, J; Mårtensson, A; Eklund, M; Libiseller, C

    2008-01-01

    In Sweden, large amounts of wood waste are generated annually from construction and demolition activities, but also from other discarded products such as packaging and furniture. A large share of this waste is today recovered and used for heat production. However, previous research has found that recovered wood waste (RWW) contains hazardous substances, which has significant implications for the environmental performance of recycling. Improved sorting is often suggested as a proper strategy to decrease such implications. In this study, we aim to analyse the impacts of waste regulation on the contamination of RWW. The occurrence of industrial preservative-treated wood, which contains several hazardous substances, was used as an indicator for contamination. First the management of RWW during 1995-2004 was studied through interviews with involved actors. We then determined the occurrence of industrial preservative-treated wood in RWW for that time period for each supplier (actor). From the results, it can be concluded that a substantially less contaminated RWW today relies on extensive source separation. The good news is that some actors, despite several obstacles for such upstream efforts, have already today proved capable of achieving relatively efficient separation. In most cases, however, the existing waste regulation has not succeeded in establishing strong enough incentives for less contaminated waste in general, nor for extensive source separation in particular. One important factor for this outcome is that the current market forces encourage involved actors to practice weak quality requirements and to rely on end-of-pipe solutions, rather than put pressure for improvements on upstream actors. Another important reason is that there is a lack of communication and oversight of existing waste regulations. Without such steering mechanisms, the inherent pressure from regulations becomes neutralized.

  8. Thermoexoemission detectors for monitoring radioactive contamination of industrial waste waters

    International Nuclear Information System (INIS)

    Obukhov, V.T.; Sobolev, I.A.; Khomchik, L.M.

    1987-01-01

    Detectors on base of BeO(Na) monocrystals with thermoemission to be used for monitoring radioactive contamination of industrial waste waters are suggested. The detectors advantages are sensitivity to α and low-ehergy β radiations, high mechanical strength and wide range of measurements. The main disadvantage is the necessity of working in red light

  9. DOE requests waiver on double containment for HLW canisters

    International Nuclear Information System (INIS)

    Lobsenz, G.

    1994-01-01

    The Energy Department has asked the Nuclear Regulatory Commission to waive double containment requirements for vitrified high-level radioactive waste canisters, saying the additional protection is not necessary and too costly. NRC said it had received a petition from DOE contending that the vitrified waste canisters were durable enough without double containment to prevent any potential plutonium release during handling and shipping. DOE said testing had shown that the vitrified waste canisters were similar - even superior - in durability to spent reactor fuel shipments, which NRC specifically exempted from the double containment requirement

  10. Grain boundary corrosion of copper canister material

    International Nuclear Information System (INIS)

    Fennell, P.A.H.; Graham, A.J.; Smart, N.R.; Sofield, C.J.

    2001-03-01

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container fitted inside a corrosion-resistant copper canister. During fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities within the copper at the new grain boundaries. The work described in this report was undertaken to determine whether there is any possibility of enhanced corrosion at grain boundaries within the copper canister. The potential for grain boundary corrosion was investigated by exposing copper specimens, which had undergone different heat treatments and hence had different grain sizes, to aerated artificial bentonite-equilibrated groundwater with two concentrations of chloride, for increasing periods of time. The degree of grain boundary corrosion was determined by atomic force microscopy (AFM) and optical microscopy. AFM showed no increase in grain boundary 'ditching' for low chloride groundwater. In high chloride groundwater the surface was covered uniformly with a fine-grained oxide. No increases in oxide thickness were observed. No significant grain boundary attack was observed using optical microscopy either. The work suggests that in aerated artificial groundwaters containing chloride ions, grain boundary corrosion of copper is unlikely to adversely affect SKB's copper canisters

  11. Oxygen incineration process for treatment of alpha-contaminated wastes

    International Nuclear Information System (INIS)

    Kim, Jeong Guk; Yang, Hee Chul; Park, Geun Il; Kim, In Tae; Kim, Joon Hyung

    2001-07-01

    As a part of development of a treatment technology for burnable alpha-bearing (or -contaminated) wastes using an oxygen incineration process, which would be expected to produce in Korea, the off-gas volume and compositions were estimated form mass and heat balance, and then compared to those of a general air incineration process. A laboratory-scale oxygen incineration process, to investigate a burnable wastes from nuclear fuel fabricatin facility, was designed, constructed, and then operated. The use of oxygen instead of air in incineratin would result in reduction on off-gas product below one seventh theoretically. In addition, the trends on incineration and melting processes to treat the radioactive alpha-contaminated wastes, and the regulations and guide lines, related to design, construction, and operation of incineration process, were reviewed. Finallu, the domestic regulations related incineration, and the operation and maintenance manuals for oxy-fuel burner and oxygen incineration process were shown in appendixes

  12. Oxygen incineration process for treatment of alpha-contaminated wastes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jeong Guk; Yang, Hee Chul; Park, Geun Il; Kim, In Tae; Kim, Joon Hyung

    2001-07-01

    As a part of development of a treatment technology for burnable alpha-bearing (or -contaminated) wastes using an oxygen incineration process, which would be expected to produce in Korea, the off-gas volume and compositions were estimated form mass and heat balance, and then compared to those of a general air incineration process. A laboratory-scale oxygen incineration process, to investigate a burnable wastes from nuclear fuel fabricatin facility, was designed, constructed, and then operated. The use of oxygen instead of air in incineratin would result in reduction on off-gas product below one seventh theoretically. In addition, the trends on incineration and melting processes to treat the radioactive alpha-contaminated wastes, and the regulations and guide lines, related to design, construction, and operation of incineration process, were reviewed. Finallu, the domestic regulations related incineration, and the operation and maintenance manuals for oxy-fuel burner and oxygen incineration process were shown in appendixes.

  13. Drop Calculations of HLW Canister and Pu Can-in-Canister

    International Nuclear Information System (INIS)

    Sreten Mastilovic

    2001-01-01

    The objective of this calculation is to determine the structural response of the standard high-level waste (HLW) canister and the canister containing the cans of immobilized plutonium (Pu) (''can-in-canister'' [CIC] throughout this document) subjected to drop DBEs (design basis events) during the handling operation. The evaluated DBE in the former case is 7-m (23-ft) vertical (flat-bottom) drop. In the latter case, two 2-ft (0.61-m) corner (oblique) drops are evaluated in addition to the 7-m vertical drop. These Pu CIC calculations are performed at three different temperatures: room temperature (RT) (20 C), T = 200 F = 93.3 C , and T = 400 F = 204 C ; in addition to these the calculation characterized by the highest maximum stress intensity is performed at T = 750 F = 399 C as well. The scope of the HLW canister calculation is limited to reporting the calculation results in terms of: stress intensity and effective plastic strain in the canister, directional residual strains at the canister outer surface, and change of canister dimensions. The scope of Pu CIC calculation is limited to reporting the calculation results in terms of stress intensity, and effective plastic strain in the canister. The information provided by the sketches from Reference 26 (Attachments 5.3,5.5,5.8, and 5.9) is that of the potential CIC design considered in this calculation, and all obtained results are valid for this design only. This calculation is associated with the Plutonium Immobilization Project and is performed by the Waste Package Design Section in accordance with Reference 24. It should be noted that the 9-m vertical drop DBE, included in Reference 24, is not included in the objective of this calculation since it did not become a waste acceptance requirement. AP-3.124, ''Calculations'', is used to perform the calculation and develop the document

  14. Mechanical integrity of canisters

    International Nuclear Information System (INIS)

    Nilsson, Fred

    1992-12-01

    This document constitutes the final report from 'SKBs reference group for mechanical integrity of canisters for spent nuclear fuel'. A complete list of all reports initiated by the reference group can be found in the summary report in this document. The main task of the reference group has been to advice SKB regarding the choice (ranking of alternatives) of canister type for different types of storage. The choice should be based on requirements of impermeability for a given time period and identification of possible limiting mechanisms. The main conclusions from the work were: From mechanical point of view, low phosphorous oxygen free copper (Cu-OFP) is a preferred canisters material. It exhibits satisfactory ductility both during tensile and creep testing. The residual stresses in the canisters are of such a magnitude that the estimated time to creep rupture with the data obtained for the Cu-OFP material is essentially infinite. Based on the present knowledge of stress corrosion cracking of copper there appears to be a small risk for such to occur in the projected environment. This risk need some further study. Rock shear movements of the size of 10 cm should pose no direct threat to the integrity of the canisters. Considering mechanical integrity, the composite copper/steel canister is an advantageous alternative. The recommendations for further research included continued studies of the creep properties of copper and of stress corrosion cracking. However, the studies should focus more directly on the design and fabrication aspect of the canister

  15. The concrete canister program

    International Nuclear Information System (INIS)

    Ohta, M.M.

    1978-02-01

    In the spring of 1974, WNRE began development and demonstration of a dry storage concept, called the concrete canister, as a possible alternative to storage of irradiated CANDU fuel in water pools. The canister is a thick-walled concrete monolith containing baskets of fuel in the dry state. The decay heat from the fuel is dissipated to the environment by natural heat transfer. Four canisters were designed and constructed. Two canisters containing electric heaters have been subjected to heat loads of 2.5 times the design, ramp heat-load cycling, and simulated weathering tests. The other two canisters were loaded with irradiated fuel, one containing fuel bundles of uniform decay heat and the other containing bundles of non-uniform decay heat in a non-symmetrical radial and axial array. The collected data were used to verify the analytical tools for prediction of effectiveness of heat transfer and radiation shielding and to verify the design of the basket and canisters. The demonstration canisters have shown that this concept is a viable alternative to water pools for the storage of irradiated CANDU fuel. (author)

  16. Can-in-canister cold demonstration in DWPF (U)

    International Nuclear Information System (INIS)

    Kuehn, N.H.

    1996-07-01

    The Department of Energy Fissile Materials Disposition Program is evaluating a number of options for disposition of weapons-usable plutonium surplus to national defense needs. One of the immobilization options is the Can-In-Canister approach. In this option small cans of a plutonium glass, which contains a neutron absorber, are placed on a support structure in a large Savannah River Site Defense Waste Processing Facility (DWPF) canister. The top is then welded onto the canister. This canister is filled with High Level Waste (HLW) glass at the DWPF. The HLW glass provides the radiation source for proliferation resistance. These canisters are to be placed in a Federal Repository. To provide information on the technical feasibility of this option prior to the Record of Decision on plutonium disposition, the Department of Energy Fissile Materials Disposition Program funded a demonstration in the DWPF. This demonstration was conducted before the start of radioactive operations. Two test canisters containing cans of surrogate (non- radioactive) plutonium glass were successfully filled with simulated HLW glass at the DWPF using standard pouring procedures. One canister had twenty cans of surrogate plutonium glass. The other had eight cans of surrogate plutonium glass. After the canisters were filled, the contents of the canisters were examined to provide data on the effect of the rack and cans on the filling of the DWPF canister, the effect of the pour on the surrogate plutonium glass and the effect of the rack and cans on the simulated HLW glass. There was no deformation of the support racks during the pour. The simulated HLW glass filled all the regions around the rack and cans and the regions between the cans and the wall of the canister. This report discusses the design of the racks and cans, the modification of the DWPF canisters to accommodate the rack and cans, the conditions during the pours and the results of the post pour analysis

  17. Mechanisms of contaminant migration from grouted waste

    International Nuclear Information System (INIS)

    Magnuson, S.O.; Yu, A.D.

    1992-01-01

    Low-level radioactive decontaminated salt solution is generated at the Savannah River Site (SRS) from the In-Tank Precipitation process. The solution is mixed with cement, slag, and fly ash, to form a grout, termed ''Saltstone'', that will be disposed in concrete vaults at the Saltstone Disposal Facility (SDF) [1]. Of the contaminants in the Saltstone, the greatest concern to SRS is the potential release of nitrate to the groundwater because of the high initial nitrate concentration (0.25 g/cm 3 ) in the Saltstone and the low Safe Drinking Water Act (SDWA) maximum contaminant level (MCL) of 44 mg/L. The SDF is designed to allow a slow, controlled release over thousands of years. This paper addresses a modeling study of nitrate migration from intact non-degraded concrete vaults in the unsaturated zone for the Radiological Performance Assessment (PA) of the SRS Saltstone Disposal Facility [3]. The PA addresses the performance requirements mandated by DOE Order 5820.2A [4

  18. EB-welding of the copper canister for the nuclear waste disposal. Final report of the development programme 1994-1997

    Energy Technology Data Exchange (ETDEWEB)

    Aalto, H. [Outokumpu Oy Poricopper, Pori (Finland)

    1998-10-01

    During 1994-1997 Posiva Oy and Outokumpu Poricopper Oy had a joint project Development of EB-welding method for massive copper canister manufacturing. The project was part of the national technology program `Weld 2000` and it was supported financially by Technology Development Centre (TEKES). The spent fuel from Finnish nuclear reactors is planned to be encapsulated in thick-walled copper canisters and placed deep into the bedrock. The thick copper layer of the canister provides a long time corrosion resistance and prevents deposited nuclear fuel from contact with water. The quality requirements of the copper components are high because of the designed long lifetime of the canister. The EB-welding technology has proved to be applicable method for the production of the copper canisters and the EB-welding technique is needed at least when the lids of the copper canister will be closed. There are a number of parameters in EB-welding which affect weldability. However, the effect of the welding parameters and their optimization has not been extensively studied in welding of thick copper sections using conventional high vacuum EB-welding. One aim of this development work was to extensively study effect of welding parameters on weld quality. The final objective was to minimise welding defects in the main weld and optimize slope out procedure in thick copper EB-welding. Welding of 50 mm thick copper sections was optimized using vertical and horizontal EB-welding techniques. As a result two full scale copper lids were welded to a short cylinder successfully. The resulting weld quality with optimised welding parameters was reasonable good. The optimised welding parameters for horizontal and vertical beam can be applied to the longitudinal body welds of the canister. The optimal slope out procedure for the lid closure needs some additional development work. In addition of extensive EB-welding program ultrasonic inspection and creep strength of the weld were studied. According

  19. EB-welding of the copper canister for the nuclear waste disposal. Final report of the development programme 1994-1997

    International Nuclear Information System (INIS)

    Aalto, H.

    1998-10-01

    During 1994-1997 Posiva Oy and Outokumpu Poricopper Oy had a joint project Development of EB-welding method for massive copper canister manufacturing. The project was part of the national technology program 'Weld 2000' and it was supported financially by Technology Development Centre (TEKES). The spent fuel from Finnish nuclear reactors is planned to be encapsulated in thick-walled copper canisters and placed deep into the bedrock. The thick copper layer of the canister provides a long time corrosion resistance and prevents deposited nuclear fuel from contact with water. The quality requirements of the copper components are high because of the designed long lifetime of the canister. The EB-welding technology has proved to be applicable method for the production of the copper canisters and the EB-welding technique is needed at least when the lids of the copper canister will be closed. There are a number of parameters in EB-welding which affect weldability. However, the effect of the welding parameters and their optimization has not been extensively studied in welding of thick copper sections using conventional high vacuum EB-welding. One aim of this development work was to extensively study effect of welding parameters on weld quality. The final objective was to minimise welding defects in the main weld and optimize slope out procedure in thick copper EB-welding. Welding of 50 mm thick copper sections was optimized using vertical and horizontal EB-welding techniques. As a result two full scale copper lids were welded to a short cylinder successfully. The resulting weld quality with optimised welding parameters was reasonable good. The optimised welding parameters for horizontal and vertical beam can be applied to the longitudinal body welds of the canister. The optimal slope out procedure for the lid closure needs some additional development work. In addition of extensive EB-welding program ultrasonic inspection and creep strength of the weld were studied. According

  20. Technological features of contamination and purification of drilling waste water

    Energy Technology Data Exchange (ETDEWEB)

    Striletskiy, I V

    1981-01-01

    The most efficient solution to the problem of preventing contamination of the reservoirs with waste water is their reuse for water supply of the borehole. Requirements are presented which the purified waste water must meet. As a result of the conducted studies it has been established that in reservoirs, only coarsely dispersed mixture, weighting compounds and floating petroleum products are removed from the water. Finely dispersed suspension and colloid particles have a sedimentation stability and do not settle out under the influence of the gravity force. For drilling waste water there is a characteristic inconsistency in the degree of contamination both at the different boreholes and at one borehole with the passage of time. Physical-chemical characteristics of the waste waters are presented. The greatest degree of contamination of water is observed when such operations are performed as replacement of the drilling fluid, lifting of the drilling tool, cementing as well as the development of emergencies. Studies on the purification of drilling water were conducted on an experimental-industrial unit.

  1. Groundwork for Universal Canister System Development

    Energy Technology Data Exchange (ETDEWEB)

    Price, Laura L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gross, Mike [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Prouty, Jeralyn L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rigali, Mark J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Craig, Brian [Argonne National Lab. (ANL), Argonne, IL (United States); Han, Zenghu [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, John Hok [Argonne National Lab. (ANL), Argonne, IL (United States); Liu, Yung [Argonne National Lab. (ANL), Argonne, IL (United States); Pope, Ron [Argonne National Lab. (ANL), Argonne, IL (United States); Connolly, Kevin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Feldman, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jarrell, Josh [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Radulescu, Georgeta [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wells, Alan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    The mission of the United States Department of Energy's Office of Environmental Management is to complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and go vernment - sponsored nuclear energy re search. S ome of the waste s that that must be managed have be en identified as good candidates for disposal in a deep borehole in crystalline rock (SNL 2014 a). In particular, wastes that can be disposed of in a small package are good candidates for this disposal concept. A canister - based system that can be used for handling these wastes during the disposition process (i.e., storage, transfers, transportation, and disposal) could facilitate the eventual disposal of these wastes. This report provides information for a program plan for developing specifications regarding a canister - based system that facilitates small waste form packaging and disposal and that is integrated with the overall efforts of the DOE's Office of Nuclear Energy Used Fuel Dis position Camp aign's Deep Borehole Field Test . Groundwork for Universal Ca nister System Development September 2015 ii W astes to be considered as candidates for the universal canister system include capsules containing cesium and strontium currently stored in pools at the Hanford Site, cesium to be processed using elutable or nonelutable resins at the Hanford Site, and calcine waste from Idaho National Laboratory. The initial emphasis will be on disposal of the cesium and strontium capsules in a deep borehole that has been drilled into crystalline rock. Specifications for a universal canister system are derived from operational, performance, and regulatory requirements for storage, transfers, transportation, and disposal of radioactive waste. Agreements between the Department of Energy and the States of Washington and Idaho, as well as the Deep Borehole Field Test plan provide schedule requirements for development of the universal canister system

  2. Physicochemical and Geotechnical Alterations to MX-80 Bentonite at the Waste Canister Interface in an Engineered Barrier System

    Directory of Open Access Journals (Sweden)

    Christopher W. Davies

    2017-08-01

    Full Text Available The study investigated the basic geomechanical and mineralogical evolution of the bentonite barrier under various experimental boundary conditions which replicated the near-field Thermo-Hydro-Chemico (THC conditions in a repository. The relationships between the physicochemical alterations and changes in the geotechnical properties have seldom been studied, especially on a consistent dataset. This paper attempts to link the physicochemical properties of Na-bentonite (MX-80 to the macro-scale engineering functionality of the bentonite post THC exposure. Experiments investigated the impact of THC variables on the engineering and physicochemical functionality of the bentonite with respect to its application within a High-Level Waste (HLW engineered barrier system. Intrinsic alterations to the MX-80 bentonite under relatively short-term exposure to hydrothermal and chemical conditions were measured. Additionally, two long-term tests were conducted under ambient conditions to consider the impact of exposure duration. The intrinsic measurements were then related to the overall performance of the bentonite as a candidate barrier material for application in a UK geological disposal facility. Findings indicate that exposure to thermo-saline-corrosion conditions (i.e., corrosion products derived from structural grade 275 carbon steel inhibits the free swell capacity and plasticity of the bentonite. However, the measured values remained above the design limits set out for the Swedish multi-barrier concept, from which the UK concept may take a lead. Corrosion alone does not appear to significantly affect the geotechnical measurements compared with the influence of thermal loading and high saline pore water after relatively short-term exposure. Thermal and corrosion exposure displayed no impact on the intrinsic swelling of the smectite component, indicating that no significant structural alteration had occurred. However, when exploring more complex saline

  3. Chemical tailoring of steam to remediate underground mixed waste contaminents

    Science.gov (United States)

    Aines, Roger D.; Udell, Kent S.; Bruton, Carol J.; Carrigan, Charles R.

    1999-01-01

    A method to simultaneously remediate mixed-waste underground contamination, such as organic liquids, metals, and radionuclides involves chemical tailoring of steam for underground injection. Gases or chemicals are injected into a high pressure steam flow being injected via one or more injection wells to contaminated soil located beyond a depth where excavation is possible. The injection of the steam with gases or chemicals mobilizes contaminants, such as metals and organics, as the steam pushes the waste through the ground toward an extraction well having subatmospheric pressure (vacuum). The steam and mobilized contaminants are drawn in a substantially horizontal direction to the extraction well and withdrawn to a treatment point above ground. The heat and boiling action of the front of the steam flow enhance the mobilizing effects of the chemical or gas additives. The method may also be utilized for immobilization of metals by using an additive in the steam which causes precipitation of the metals into clusters large enough to limit their future migration, while removing any organic contaminants.

  4. Hanford tank residual waste - Contaminant source terms and release models

    International Nuclear Information System (INIS)

    Deutsch, William J.; Cantrell, Kirk J.; Krupka, Kenneth M.; Lindberg, Michael L.; Jeffery Serne, R.

    2011-01-01

    Highlights: → Residual waste from five Hanford spent fuel process storage tanks was evaluated. → Gibbsite is a common mineral in tanks with high Al concentrations. → Non-crystalline U-Na-C-O-P ± H phases are common in the U-rich residual. → Iron oxides/hydroxides have been identified in all residual waste samples. → Uranium release is highly dependent on waste and leachant compositions. - Abstract: Residual waste is expected to be left in 177 underground storage tanks after closure at the US Department of Energy's Hanford Site in Washington State, USA. In the long term, the residual wastes may represent a potential source of contamination to the subsurface environment. Residual materials that cannot be completely removed during the tank closure process are being studied to identify and characterize the solid phases and estimate the release of contaminants from these solids to water that might enter the closed tanks in the future. As of the end of 2009, residual waste from five tanks has been evaluated. Residual wastes from adjacent tanks C-202 and C-203 have high U concentrations of 24 and 59 wt.%, respectively, while residual wastes from nearby tanks C-103 and C-106 have low U concentrations of 0.4 and 0.03 wt.%, respectively. Aluminum concentrations are high (8.2-29.1 wt.%) in some tanks (C-103, C-106, and S-112) and relatively low ( 2 -saturated solution, or a CaCO 3 -saturated water. Uranium release concentrations are highly dependent on waste and leachant compositions with dissolved U concentrations one or two orders of magnitude higher in the tests with high U residual wastes, and also higher when leached with the CaCO 3 -saturated solution than with the Ca(OH) 2 -saturated solution. Technetium leachability is not as strongly dependent on the concentration of Tc in the waste, and it appears to be slightly more leachable by the Ca(OH) 2 -saturated solution than by the CaCO 3 -saturated solution. In general, Tc is much less leachable (<10 wt.% of the

  5. Report on the state of radiation contamination in disaster waste

    International Nuclear Information System (INIS)

    Onishi, Yuko; Sasaki, Satoru; Yamada, Norikazu; Kawasaki, Satoru

    2011-09-01

    Fukushima Prefecture faces extreme difficulties in disposing of waste generated from the tsunami disaster (hereinafter referred to as disaster waste) and contaminated with radioactive material released from the crippled Fukushima Dai-ichi Nuclear Power Station. Although the waste should be treated according to the level of radioactivity, there are only air dose rates and radionuclide analyses of soil due to monitoring around the Fukushima Dai-ichi Nuclear Power Station and there has been no information on the radioactivity concentration of the disaster waste. The radioactivity concentration of the disaster waste was investigated by sampling measurement and in-situ Ge measurement at 20 temporary disaster waste storages in Fukushima Prefecture excluding the evacuation zone and 'deliberate evacuation zone.' JNES carried out this investigation upon a request from the Nuclear and Industrial Safety Agency. The investigation revealed that the measured radioactivity concentrations of the disaster waste lumps were enveloped within the soil monitoring readings in Fukushima Prefecture and also within a correlated curve between the air dose rates obtained from air dose rate readings around the disaster waste and the radioactivity concentrations of it. With these correlation curves, the radioactivity concentration of the disaster waste is estimated to be less than 8,000 Bq/kg at almost all places in the affected area excluding the evacuation zone and 'deliberate evacuation zone.' Measurements by in-situ Ge showed that the radioactivity of the disaster waste which had been expected to be more than 8,000 Bq/kg was less than 8,000 Bq/kg. (author)

  6. Release of low-contaminated reactor wastes for unrestricted use

    International Nuclear Information System (INIS)

    Carleson, G.

    1982-01-01

    A generic methodology has been used to evaluate the dose contributions to an individual and to the population of five categories of low-contaminated reactor wastes produced according to the Swedish program and released for unrestricted handling and use. A reference quantity with a surface dose rate below a predetermined level is followed along the whole commercial pathway from the reactor station to the final product consumer and/or a municipal waste station. Dose contributions are calculated for each step in a normal pathway under maximally unfavourable conditions. (Auth.)

  7. Solid-waste leach characteristics and contaminant-sediment interactions

    International Nuclear Information System (INIS)

    Serne, R.J.; LeGore, V.L.; Cantrell, K.J.; Lindenmeier, C.W.; Campbell, J.A.; Amonette, J.E.; Conca, J.L.; Wood, M.I.

    1993-10-01

    The objectives of this report and subsequent volumes include describing progress on (1) development of conceptual-release models for Hanford Site defense solid-waste forms; (2) optimization of experimental methods to quantify the release from contaminants from solid wastes and their subsequent interactions with unsaturated sediments; and (3) creation of empirical data for use as provisional source term and retardation factors that become input parameters for performance assessment analyses for future Hanford disposal units and baseline risk assessments for inactive and existing disposal units

  8. Beta-gamma contaminated solid waste incinerator facility

    International Nuclear Information System (INIS)

    Hootman, H.E.

    1979-10-01

    This technical data summary outlines a reference process to provide a 2-stage, 400 lb/hour incinerator to reduce the storage volume of combustible process waste contaminated with low-level beta-gamma emitters in response to DOE Manual 0511. This waste, amounting to more than 200,000 ft 3 per year, is presently buried in trenches in the burial ground. The anticipated storage volume reduction from incineration will be a factor of 20. The incinerator will also dispose of 150,000 gallons of degraded solvent from the chemical separations areas and 5000 gallons per year of miscellaneous nonradioactive solvents which are presently being drummed for storage

  9. Multi-purpose canister project overview

    International Nuclear Information System (INIS)

    Williams, J.

    1995-01-01

    In this presentation, the author lists the approved and proposed dry storage technologies. He discusses the compatibility of dry storage systems with waste management systems. Historical aspects, recent history, key features of the program approach, benefits, specifications, acquisition and potential utility use of the multi-purpose canister (MPC) are covered. The MPCs provide standardization in the waste management system and a cost savings to utilities and government. MPC will be developed to the same level as existing dry storage systems

  10. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-04-07

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for

  11. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the CHF and may not reflect the ongoing design evolution of the facility

  12. Design considerations for incineration of transuranic-contaminated solid wastes

    International Nuclear Information System (INIS)

    Koenig, R.A.

    1977-01-01

    The Los Alamos Scientific Laboratory has established a development program to evaluate alternate production-level (100-200 lb/hr throughput) volume reduction processes for transuranic-contaminated solid waste. The first process selected for installation and study is based on controlled-air incineration. Design considerations leading to selection of feed preparation, incineration, residue removal, and off-gas cleanup components and their respective radioactive containment provisions will be presented

  13. Geostatistical methodology for waste optimization of contaminated premises - 59344

    International Nuclear Information System (INIS)

    Desnoyers, Yvon; Dubot, Didier

    2012-01-01

    The presented methodological study illustrates a Geo-statistical approach suitable for radiological evaluation in nuclear premises. The waste characterization is mainly focused on floor concrete surfaces. By modeling the spatial continuity of activities, Geo-statistics provide sound methods to estimate and map radiological activities, together with their uncertainty. The multivariate approach allows the integration of numerous surface radiation measurements in order to improve the estimation of activity levels from concrete samples. This way, a sequential and iterative investigation strategy proves to be relevant to fulfill the different evaluation objectives. Waste characterization is performed on risk maps rather than on direct interpolation maps (due to bias of the selection on kriging results). The use of several estimation supports (punctual, 1 m 2 , room) allows a relevant radiological waste categorization thanks to cost-benefit analysis according to the risk of exceeding a given activity threshold. Global results, mainly total activity, are similarly quantified to precociously lead the waste management for the dismantling and decommissioning project. This paper recalled the geo-statistics principles and demonstrated how this methodology provides innovative tools for the radiological evaluation of contaminated premises. The relevance of this approach relies on the presence of a spatial continuity for radiological contamination. In this case, geo-statistics provides reliable activity estimates, uncertainty quantification and risk analysis, which are essential decision-making tools for decommissioning and dismantling projects of nuclear installations. Waste characterization is then performed taking all relevant information into account: historical knowledge, surface measurements and samples. Thanks to the multivariate processing, the different investigation stages can be rationalized as regards quantity and positioning. Waste characterization is finally

  14. Waste processing system for product contaminated with radioactivity

    International Nuclear Information System (INIS)

    Sotoyama, Koichi; Takaya, Jun-ichi; Takahashi, Suehiro.

    1987-01-01

    Purpose: To enable to processing contaminated products while separating them into metals at high contamination level and non-metals at low contamination level. Constitution: Pulverized radioactive wastes conveyed on a conveyor belt are uniformly irradiated by a ring-illumination device and then they are picked-up by a television camera or the like. The picked-up signals are sent to an image processing device, applied with appropriate binarization and metal objects are separated by utilizing the light absorbing property of non-metal and light reflection property of metals. The graviational center for the metal object is calculated from the binarized image, the positional information is provided to a robot controller and the metal object is transferred to another position by a robot. Since only the metal object at high radioactive contamination level can be taken out separately, it is no more necessary to process the entire wastes as the high level decontamination products, to thereby provide an economical advantage. (Sekiya, K.)

  15. Sanitation of conditioned radioactive waste after a contamination accident

    International Nuclear Information System (INIS)

    Aeppli, J.

    1980-01-01

    In June 1978, there occurred in the port of Ijmuiden, Netherlands, a contamination incident involving drums originating from Switzerkand and containing radioactive wastes intended to be dumped into the sea. The batch of 207 drums excluded from the sea-dumping action had to be sanitated for the next year dumping in such a manner, that these wastes met the international requirements and could be disposed of by sinking them into the Atlantic. As a consequence of extensive sanitation work, requiring part of the wastes to be newly conditioned and several drums to be packaged again, the total weight of the wastes ready for dumping was doubled. The total radiation exposure for the personnel that took part in the individual phases of sanitation amounted to about 10 man-rem. The main causes for this contamination incident were unusual chemical composition of the concentrate to be solidified, unsufficient quality control and a position not suitabble for transport. The measures taken after this incident intend to avoid similar occurrences in the future. (orig.) [de

  16. Radioactive waste and contamination in the former Soviet Union

    International Nuclear Information System (INIS)

    Suokko, K.; Reicher, D.

    1993-01-01

    Decades of disregard for the hazards of radioactive waste have created contamination problems throughout the former Soviet Union rivaled only by the Chernobyl disaster. Although many civilian activities have contributed to radioactive waste problems, the nuclear weapons program has been by far the greatest culprit. For decades, three major weapons production facilities located east of the Ural Mountains operated in complete secrecy and outside of environmental controls. Referred to until recently only by their postal abbreviations, the cities of Chelyabinsk-65, Tomsk-7, and Krasnoyarsk-26 were open only to people who worked in them. The mismanagement of waste at these sites has led to catastrophic accidents and serious releases of radioactive materials. Lack of public disclosure, meanwhile, has often prevented proper medical treatment and caused delays in cleanup and containment. 5 refs

  17. Disposal of slightly contaminated radioactive wastes from nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Minns, J.L. [Nuclear Regulatory Commission, Washington, DC (United States)

    1995-02-01

    With regard to the disposal of solid wastes, nuclear power plants basically have two options, disposal in a Part 61 licensed low-level waste site, or receive approval pursuant to 20.2002 for disposal in a manner not otherwise authorized by the NRC. Since 1981, the staff has reviewed and approved 30 requests for disposal of slightly contaminated radioactive materials pursuant to Section 20.2002 (formerly 20.302) for nuclear power plants located in non-Agreement States. NRC Agreement States have been delegated the authority for reviewing and approving such disposals (whether onsite or offsite) for nuclear power plants within their borders. This paper describes the characteristics of the waste disposed of, the review process, and the staff`s guidelines.

  18. Feasibility of disposal of high-level radioactive waste into the seabed. Volume 8: Review of processes near a buried waste canister

    International Nuclear Information System (INIS)

    Lanza, F.

    1988-01-01

    One of the options suggested for disposal of high-level radioactive waste resulting from the generation of nuclear power is burial beneath the deep ocean floor in geologically stable sediment formations which have no economic value. The 8-volume series provides an assessment of the technical feasibility and radiological safety of this disposal concept based on the results obtained by ten years of co-operation and infomation exchange among the Member countries participating in the NEA Seabed Working Group. This report investigates the phenomena arriving in the proximity of the waste package immersed in the sea sediments

  19. Process for affixing radioactive contamination on contaminated materials or wastes. Its application

    International Nuclear Information System (INIS)

    Aubouin, Guy; Aude, Georges; Tassigny, Christian de.

    1982-01-01

    The invention concerns a process for affixing radioactive contamination on materials or waste matter in order to ensure that the materials are transferred in complete safety or to package them when their activity is low. Under this process at least one coat of a resin polymerizable at ambient temperature, for example an epoxy resin, a polyester resin, a vinyl resin or a mixture of thermohardening resin and thermoplastic resin, is sprayed on to the contaminated material part by means of an electrostatic gun [fr

  20. Manufacture of disposal canisters

    International Nuclear Information System (INIS)

    Nolvi, L.

    2009-12-01

    The report summarizes the development work carried out in the manufacturing of disposal canister components, and present status, in readiness for manufacturing, of the components for use in assembly of spent nuclear fuel disposal canister. The disposal canister consist of two major components: the nodular graphite cast iron insert and overpack of oxygen-free copper. The manufacturing process for copper components begins with a cylindrical cast copper billet. Three different manufacturing processes i.e. pierce and draw, extrusion and forging are being developed, which produce a seamless copper tube or a tube with an integrated bottom. The pierce and draw process, Posiva's reference method, makes an integrated bottom possible and only the lid requires welding. Inserts for BWR-element are cast with 12 square channels and inserts for VVER 440-element with 12 round channels. Inserts for EPR-elements have four square channels. Casting of BWR insert type has been studied so far. Experience of casting inserts for PWR, which is similar to the EPR-type, has been got in co-operation with SKB. The report describes the processes being developed for manufacture of disposal canister components and some results of the manufacturing experiments are presented. Quality assurance and quality control in manufacture of canister component is described. (orig.)

  1. SOURCE TERMS FOR HLW GLASS CANISTERS

    International Nuclear Information System (INIS)

    J.S. Tang

    2000-01-01

    This calculation is prepared by the Monitored Geologic Repository (MGR) Waste Package Design Section. The objective of this calculation is to determine the source terms that include radionuclide inventory, decay heat, and radiation sources due to gamma rays and neutrons for the high-level radioactive waste (HLW) from the, West Valley Demonstration Project (WVDP), Savannah River Site (SRS), Hanford Site (HS), and Idaho National Engineering and Environmental Laboratory (INEEL). This calculation also determines the source terms of the canister containing the SRS HLW glass and immobilized plutonium. The scope of this calculation is limited to source terms for a time period out to one million years. The results of this calculation may be used to carry out performance assessment of the potential repository and to evaluate radiation environments surrounding the waste packages (WPs). This calculation was performed in accordance with the Development Plan ''Source Terms for HLW Glass Canisters'' (Ref. 7.24)

  2. Plutonium contaminated solid waste programs at the Los Alamos Scientific Laboratory

    International Nuclear Information System (INIS)

    Johson, L.J.; Jordan, H.S.

    1975-01-01

    Development of handling and storage criteria for plutonium contaminated solid waste materials is discussed. Data from corrosion and radiolytic attack studies are reviewed. Instrumentation systems developed for solid waste management applications at the 10nCi Pu/g waste material level is described and their sensitivity and operational limitations reviewed. Current programs for the environmental risk analysis of past waste disposal areas and for development of technology for the volume reduction and chemical stabilization of transuranic contaminated solid waste is outlined

  3. Current technics and management strategy for Pu-contaminated wastes at PNC

    International Nuclear Information System (INIS)

    1981-02-01

    Power Reactor and Nuclear Fuel Development Corporation (PNC) was designated as a leading organization for the Pu-contaminated waste technology program in Japan. For this purpose, number of efforts in the research and development are proceeding. That is, Pu-contaminated waste technology including volume reduction system and the immobilization of wastes is being developed. The design of a Pu-contaminated waste treatment facility (PWTF) is being made for the demonstration of the technology developed. Studies are in progress to find the criteria for waste products in disposal. The current procedures and strategy for the management of Pu-contaminated wastes at PNC are described as follows: current and future management; technology development including controlled air incineration, acid digestion, immobilization melting, dismantling, and liquid waste treatment; the Pu-contaminated waste treatment facility. (J.P.N.)

  4. Environmental assessment of waste matrices contaminated with arsenic.

    Science.gov (United States)

    Sanchez, F; Garrabrants, A C; Vandecasteele, C; Moszkowicz, P; Kosson, D S

    2003-01-31

    The use of equilibrium-based and mass transfer-based leaching tests has been proposed to provide an integrated assessment of leaching processes from solid wastes. The objectives of the research presented here are to (i) validate this assessment approach for contaminated soils and cement-based matrices, (ii) evaluate the use of diffusion and coupled dissolution-diffusion models for estimating constituent release, and (iii) evaluate model parameterization using results from batch equilibrium leaching tests and physical characterization. The test matrices consisted of (i) a soil contaminated with arsenic from a pesticide production facility, (ii) the same soil subsequently treated by a Portland cement stabilization/solidification (S/S) process, and (iii) a synthetic cement-based matrix spiked with arsenic(III) oxide. Results indicated that a good assessment of contaminant release from contaminated soils and cement-based S/S treated wastes can be obtained by the integrated use of equilibrium-based and mass transfer-based leaching tests in conjunction with the appropriate release model. During the time scale of laboratory testing, the release of arsenic from the contaminated soil matrix was governed by diffusion and the solubility of arsenic in the pore solution while the release of arsenic from the cement-based matrices was mainly controlled by solubilization at the interface between the matrix and the bulk leaching solution. In addition, results indicated that (i) estimation of the activity coefficient within the matrix pore water is necessary for accurate prediction of constituent release rates and (ii) inaccurate representation of the factors controlling release during laboratory testing can result in significant errors in release estimates.

  5. Assay of plutonium contaminated waste by gamma spectrometry

    International Nuclear Information System (INIS)

    Adsley, I.; Bull, R.; Davies, M.; Green, M.

    2011-01-01

    The extreme toxicity of plutonium necessitates the segregation of plutonium contaminated materials (PCM) with extremely small (sub-μg) levels of contamination. The driver to measure accurately these small quantities of plutonium within (relatively) large volumes of waste is (in part) financial. In particular the cost of disposal (per unit volume) rises steeply with increasing waste-category. Within the UK, there has been a historical reluctance to use low energy gamma radiation to sentence PCM because of the potential for self attenuation by dense materials. This is unfortunate because the low-energy gamma radiation from PCM offers the only practicable technique for segregating PCM within the various Low Level Waste (LLW) (>0.4Bq/g) and sub-LLW categories. Whilst passive neutron counting techniques have proved successful for assay of waste well into the Intermediate Level Waste (ILW) (>100Bq/g) category, a cursory study reveals that these techniques are barely capable of detecting mg quantities of plutonium -- let alone the sub-μg quantities present in LLW. This paper considers the use of two types of gamma detector for assay of PCM: the thin sodium iodide FIDLER (Field Instrument for the Detection of Low Energy Radiation) and the HPGe (High Purity Germanium) detector. Systems utilising these two types of detector can provide complementary information. FIDLER measurements are conducted by careful, local, systematic monitoring of surfaces. By contrast a HPGe detector can be used to monitor entire walls, or even rooms, in one measurement. Thus, a HPGe detector placed in the centre of room (from which any radioactive hot-spots have previously been removed) could be used to demonstrate that the average activity remaining close to the surface of the walls/floor/ceiling is below a given limit. The Monte Carlo Code MCNP 1 has been used to model both FIDLER probe and HPGe detector in the measurement geometries described above. The MCNP simulations have been validated

  6. A natural analogue for copper waste canisters: The copper-uranium mineralised concretions in the Permian mudrocks of south Devon, United Kingdom

    Energy Technology Data Exchange (ETDEWEB)

    Milodowski, A.E.; Styles, M.T.; Hards, V.L. [Natural Environment Research Council (United Kingdom). British Geological Survey

    2000-08-01

    This report presents the results of a small-scale pilot study of the mineralogy and alteration characteristics of unusual sheet-like native copper occurring together with uraniferous and vanadiferous concretions in mudstones and siltstones of the Permian Littleham Mudstone Formation, at Littleham Cove, south Devon, England. The host mudstones and siltstones are smectitic and have been compacted through deep Mesozoic burial. The occurrence of native copper within these rocks represents a natural analogue for the long-term behaviour of copper canisters, sealed in a compacted clay (bentonite) backfill, that will be used for the deep geological disposal of high-level radioactive waste by the SKB. The study was undertaken by the British Geological Survey (BGS) on behalf of SKB between November 1999 and June 2000. The study was based primarily on archived reference material collected by the BGS during regional geological and mineralogical surveys of the area in the 1970's and 1980's. However, a brief visit was made to Littleham Cove in January 2000 to try to examine the native copper in situ and to collect additional material. Unfortunately, recent landslips and mudflows obscured much of the outcrop, and only one new sample of native copper could be collected. The native copper occurs as thin plates, up to 160 mm in diameter, which occur parallel to bedding in the Permian Littleham Mudstone Formation at Littleham Cove (near Budleigh Salterton) in south Devon. Each plate is made up of composite stacks of individual thin copper sheets each 1-2 mm thick. The copper is very pure (>99.4% Cu) but is accompanied by minor amounts of native silver (also pure - >99%) which occurs as small inclusions within the native copper. Detailed mineralogical and petrological studies of the native copper sheets, using optical petrography, backscattered scanning electron microscopy, X-ray diffraction analysis and electron probe microanalytical techniques, reveal a complex history of

  7. A natural analogue for copper waste canisters: The copper-uranium mineralised concretions in the Permian mudrocks of south Devon, United Kingdom

    International Nuclear Information System (INIS)

    Milodowski, A.E.; Styles, M.T.; Hards, V.L.

    2000-08-01

    This report presents the results of a small-scale pilot study of the mineralogy and alteration characteristics of unusual sheet-like native copper occurring together with uraniferous and vanadiferous concretions in mudstones and siltstones of the Permian Littleham Mudstone Formation, at Littleham Cove, south Devon, England. The host mudstones and siltstones are smectitic and have been compacted through deep Mesozoic burial. The occurrence of native copper within these rocks represents a natural analogue for the long-term behaviour of copper canisters, sealed in a compacted clay (bentonite) backfill, that will be used for the deep geological disposal of high-level radioactive waste by the SKB. The study was undertaken by the British Geological Survey (BGS) on behalf of SKB between November 1999 and June 2000. The study was based primarily on archived reference material collected by the BGS during regional geological and mineralogical surveys of the area in the 1970's and 1980's. However, a brief visit was made to Littleham Cove in January 2000 to try to examine the native copper in situ and to collect additional material. Unfortunately, recent landslips and mudflows obscured much of the outcrop, and only one new sample of native copper could be collected. The native copper occurs as thin plates, up to 160 mm in diameter, which occur parallel to bedding in the Permian Littleham Mudstone Formation at Littleham Cove (near Budleigh Salterton) in south Devon. Each plate is made up of composite stacks of individual thin copper sheets each 1-2 mm thick. The copper is very pure (>99.4% Cu) but is accompanied by minor amounts of native silver (also pure - >99%) which occurs as small inclusions within the native copper. Detailed mineralogical and petrological studies of the native copper sheets, using optical petrography, backscattered scanning electron microscopy, X-ray diffraction analysis and electron probe microanalytical techniques, reveal a complex history of

  8. Contaminant Release from Residual Waste in Closed Single-Shell Tanks and Other Waste Forms Associated with the Tanks

    International Nuclear Information System (INIS)

    Deutsch, William J.

    2008-01-01

    This chapter describes the release of contaminants from the various waste forms that are anticipated to be associated with closure of the single-shell tanks. These waste forms include residual sludge or saltcake that will remain in the tanks after waste retrieval. Other waste forms include engineered glass and cementitious materials as well as contaminated soil impacted by previous tank leaks. This chapter also describes laboratory testing to quantify contaminant release and how the release data are used in performance/risk assessments for the tank waste management units and the onsite waste disposal facilities. The chapter ends with a discussion of the surprises and lessons learned to date from the testing of waste materials and the development of contaminant release models

  9. Plutonium Immobilization Project - Robotic canister loading

    International Nuclear Information System (INIS)

    Hamilton, R.L.

    2000-01-01

    The Plutonium Immobilization Program (PIP) is a joint venture between the Savannah River Site (SRS), Lawrence Livermore National Laboratory (LLNL), Argonne National Laboratory (ANL), and Pacific Northwest National Laboratory (PNNL). When operational in 2008, the PIP will fulfill the nation's nonproliferation commitment by placing surplus weapons-grade plutonium in a permanently stable ceramic form and making it unattractive for reuse. Since there are significant radiation and security concerns, the program team is developing novel and unique technology to remotely perform plutonium immobilization tasks. The remote task covered in this paper employs a jointed arm robot to load seven 3.5 inch diameter, 135-pound cylinders (magazines) through the 4 inch diameter neck of a stainless steel canister. Working through the narrow canister neck, the robot secures the magazines into a specially designed rack pre-installed in the canister. To provide the deterrent effect, the canisters are filled with a mixture of high-level waste and glass at the Defense Waste Processing Facility (DWPF)

  10. Waste Contaminants at Military Bases Working Group report

    International Nuclear Information System (INIS)

    1993-01-01

    The Waste Contaminants at Military Bases Working Group has screened six prospective demonstration projects for consideration by the Federal Advisory Committee to Develop On-Site Innovative Technologies (DOIT). These projects include the Kirtland Air Force Base Demonstration Project, the March Air Force Base Demonstration Project, the McClellan Air Force Base Demonstration Project, the Williams Air Force Base Demonstration Project, and two demonstration projects under the Air Force Center for Environmental Excellence. A seventh project (Port Hueneme Naval Construction Battalion Center) was added to list of prospective demonstrations after the September 1993 Working Group Meeting. This demonstration project has not been screened by the working group. Two additional Air Force remediation programs are also under consideration and are described in Section 6 of this document. The following information on prospective demonstrations was collected by the Waste Contaminants at Military Bases Working Group to assist the DOIT Committee in making Phase 1 Demonstration Project recommendations. The remainder of this report is organized into seven sections: Work Group Charter's mission and vision; contamination problems, current technology limitations, and institutional and regulatory barriers to technology development and commercialization, and work force issues; screening process for initial Phase 1 demonstration technologies and sites; demonstration descriptions -- good matches;demonstration descriptions -- close matches; additional candidate demonstration projects; and next steps

  11. Treatment of solid waste highly contaminated by alpha emitters

    International Nuclear Information System (INIS)

    Madic, C.; Breschet, C.; Vigreaux, B.

    1990-01-01

    In the recent years, efforts have been made in order to reduce the amount of alpha emitters essentially plutonium isotopes present in the solid wastes produced either during research experiments on fuel reprocessing, done in the Radiochemistry building in the centre d'etudes nuclearires de FONTENAY-AUX-ROSES (CEA, FRANCE), or in the MARCOULE reprocessing plant (COGEMA, FRANCE). The goals defined for the treatments of these different wastes were: to reduce their α and β, γ, contamination levels. and to recover the plutonium, an highly valuable material, and to minimize its quantity to be discharged with the wastes. To achieve these goals leaching processes using electrogenerated Ag (II (a very aggressive agent for PuO 2 )) in nitric acid solutions, were developed and several facilities were designed and built to operate the processes: ELISE and PROLIXE facilities: PILOT ASHES FACILITY for delete, the treatment of plutonium contaminated ashes (COGEMA, MARCOULE). A brief description of the process and of the different facilities will be presented in this paper; the main results obtained in ELISE and PROLIXE are also summarized

  12. The management of plutonium (alpha) contaminated waste materials (PCM)

    International Nuclear Information System (INIS)

    Sills, R.J.

    1984-01-01

    This article reviews the management strategies for plutonium contaminated materials (PCM), the techniques which have been used and developed for their implementation and what can be expected for the immediate future. In general reference is made to the situation in the U.K., but where appropriate the International context is noted. In the context of the article plutonium often occurs with other alpha-active materials and the two terms are used virtually synonymously. The technology which is described, and which is the result of substantial research and development programmes, has largely been developed with the objective of recovering the majority of plutonium prior to ultimate disposal of the waste. There is no doubt that this removal to low levels of contamination is technically feasible; indeed there are a number of methods to choose from each with its own advantages and disadvantages. The emphasis has shifted recently from the development and demonstration of technology for waste handling, treatment and disposal (although these are very important), to the assessment of the effects--social, technological and economic--of the various options available for dealing with the waste. The process is thus, one of achieving the lowest overall 'cost' to society; where 'cost' is in the broadest sense of effect on society and not in merely strict financial terms

  13. Latest movements associated with radioactive contamination and disaster waste management

    International Nuclear Information System (INIS)

    Omura, Tomomi

    2012-01-01

    As for the radioactive contamination countermeasures taken for the accident of the Fukushima Daiichi Nuclear Station of Tokyo Electric Power Company, this paper introduces in the digest version the following movements from early March to early April 2012. (1) Organizational structure. Inauguration of Nuclear Regulatory Agency, and the organizational structure of Fukushima Environment Regeneration Office of the Ministry of the Environment. (2) The Act on Special Measures concerning the Handling of Radioactive Pollution. Publication by the Ministry of the Environment on decontamination plan for three municipalities belonging to Special Decontamination Area, decontamination plan for Intensive Contamination Survey Area, new construction of disposal sites for designated waste with the level exceeding 8,000 Bq / kg, and disaster waste direct treatment project and substitute treatment project in Fukushima Prefecture. (3) Radiation exposure countermeasures. Lawmaker-initiated registration plan by Democratic Party, Liberal Democratic Party, and New Komeito. (4) Technological evaluation. Publication of the results of Decontamination Technology Demonstration Test Projects by the Cabinet Office, the Ministry of the Environment, and Fukushima Prefecture. (5) Monitoring. Full-scale implementation of radioactivity monitoring plan in Tokyo Bay in Fiscal 2012. (6) Disaster waste countermeasures. Request of the government to the local governments on the wide-area treatment of wreckage, active request to the Cement Association in cooperation with the treatment of wreckage, and positive replies from of 22 prefectures / cities regarding the acceptance of wide-area wreckage treatment. (O.A.)

  14. Bonding of radioactive contamination. III. Auger electron spectroscopic investigation

    International Nuclear Information System (INIS)

    Rankin, W.N.; Whitkop, P.G.

    1983-01-01

    The mechanisms by which radioactive contamination would be bonded to a DWPF canister surface are being investigated. Tests with low pressure water and air-injected water decontamination of radioactive specimens showed that bonding of contamination increases rapidly with postoxidation temperature. Even with the least severe temperature conditions expected on the waste glass canister, bonding is so great that decontamination cannot be affected by water-only techniques. A preoxidation film increased rather than decreased bonding. This memorandum describes detailed surface analyses of coupons simulating DWPF canister surfaces. Based on this examination we conclude: contamination will be dispersed throughout the oxide film on DWPF canisters. Contamination is concentrated at the surface, decreasing farther into the oxide film; some samples contain sludge contamination at the steel/oxide interface. This was not the case for semi-volatile (Cs 2 O) contamination; in samples with contamination at the steel/oxide interface, at least 80% of the contamination is usually in the oxide layer; no difference in contamination dispersion between preoxidized and non-preoxidized samples was found; and postoxidation atmosphere had no effect on the contamination dispersion within the oxide layer. 6 references, 9 figures

  15. Mechanical design of the storage tubes in the HWVP canister storage building

    International Nuclear Information System (INIS)

    Divona, C.J.; Fages, R.; Janicek, G.P.; Mullally, J.A.

    1993-01-01

    Canisters of high-level waste from the Hanford Waste Vitrification Plant (HWVP) will be stored in an adjacent facility, the Canister Storage Building (CSB). The canisters are stored vertically in an array of tubes within the shielded vault area of the CSB. This paper describes the mechanical design of the storage tubes, the shield floor plugs that confine the waste within the tubes and the impact absorber system used to assure that the canisters are not breached in the event of an accidental drop. Installation and testing of the components is also discussed

  16. Design basis for the copper/steel canister

    International Nuclear Information System (INIS)

    Bowyer, W.H.

    1996-02-01

    The development of the copper/iron canister which has been proposed by SKB for the containment of high level nuclear waste has been studied from the point of view of choice of materials, manufacturing technology and quality assurance. This report describes the observations on progress which have been made between March 1995 and Feb 1996 and the result of further literature studies. A first trial canister has been produced using a fabricated steel liner and an extruded copper tubular, a second one using a fabricated tubular is at an advanced stage. A change from a fabricated steel inner canister to a proposed cast canister has been justified by a criticality argument but the technology for producing a cast canister is at present untried. The microstructure achieved in the extruded copper tubular for the first canister is unacceptable. Similar problems exist with plate used for the fabricated tubular, but some more favourable structures have been achieved already by this route. Seam welding of the first tubular failed through a suspected material problem. The second fabricated tubular welded without difficulty. Welding of lids and bottoms to the copper canister is problematical.There is as yet no satisfactory non destructive test procedures for the parent metal or the welds in the copper canister material, partly due to the coarse grain size which arise in the proposed material processed by the proposed routes. Further studies are also required on crevice corrosion, galvanic attack and stress corrosion cracking in the copper 50 ppm phosphorus alloy. 28 refs

  17. MINE WASTE TECHNOLOGY PROGRAM; PHOSPHATE STABILIZATION OF HEAVY METALS CONTAMINATED MINE WASTE YARD SOILS, JOPLIN, MISSOURI NPL SITE

    Science.gov (United States)

    This document summarizes the results of Mine Waste Technology Project 22-Phosphate Stabilization of Heavy Metals-Contaminated Mine Waste Yard Soils. Mining, milling, and smelting of ores near Joplin, Missouri, have resulted in heavy metal contamination of the area. The Joplin s...

  18. Nitrification of highly contaminated waste water with retention of biomass

    International Nuclear Information System (INIS)

    Weichgrebe, D.

    1992-09-01

    The AIF Research Project No 7698 was concerned with the nitrification of highly contaminated waste water with retention of biomass. A compact system for the nitrification was developed and optimized in the investigations. This is an over-dammed fixed bed reactor with structured packing elements and membrane gasification. The fixed bed reactor was successfully installed in a multi-stage compact plant on the laboratory scale for the biological treatment of dump trickled water. With the conclusion of the investigations, design data are available for the technical scale realisation of nitrification in fixed bed reactors. (orig.) [de

  19. Characterisation of Plasma Vitrified Simulant Plutonium Contaminated Material Waste

    International Nuclear Information System (INIS)

    Hyatt, Neil C.; Morgan, Suzy; Stennett, Martin C.; Scales, Charlie R.; Deegan, David

    2007-01-01

    The potential of plasma vitrification for the treatment of a simulant Plutonium Contaminated Material (PCM) was investigated. It was demonstrated that the PuO 2 simulant, CeO 2 , could be vitrified in the amorphous calcium iron aluminosilicate component of the product slag with simultaneous destruction of the organic and polymer waste fractions. Product Consistency Tests conducted at 90 deg. C in de-ionised water and buffered pH 11 solution show the PCM slag product to be durable with respect to release of Ce. (authors)

  20. A Film Canister Colorimeter.

    Science.gov (United States)

    Gordon, James; James, Alan; Harman, Stephanie; Weiss, Kristen

    2002-01-01

    A low-cost, low-tech colorimeter was constructed from a film canister. The student-constructed colorimeter was used to show the Beer-Lambert relationship between absorbance and concentration and to calculate the value of the molar absorptivity for permanganate at the wavelength emission maximum for an LED. Makes comparisons between this instrument…

  1. Toxicity Assessment of Contaminated Soils of Solid Domestic Waste Landfill

    Science.gov (United States)

    Pasko, O. A.; Mochalova, T. N.

    2014-08-01

    The paper delivers the analysis of an 18-year dynamic pattern of land pollutants concentration in the soils of a solid domestic waste landfill. It also presents the composition of the contaminated soils from different areas of the waste landfill during its operating period. The authors calculate the concentrations of the following pollutants: chrome, nickel, tin, vanadium, lead, cuprum, zinc, cobalt, beryllium, barium, yttrium, cadmium, arsenic, germanium, nitrate ions and petrochemicals and determine a consistent pattern of their spatial distribution within the waste landfill area as well as the dynamic pattern of their concentration. Test-objects are used in experiments to make an integral assessment of the polluted soil's impact on living organisms. It was discovered that the soil samples of an animal burial site are characterized by acute toxicity while the area of open waste dumping is the most dangerous in terms of a number of pollutants. This contradiction can be attributed to the synergetic effect of the polluted soil, which accounts for the regularities described by other researchers.

  2. Controlled-air incineration of transuranic-contaminated solid waste

    International Nuclear Information System (INIS)

    Borduin, L.C.; Draper, W.E.; Koenig, R.A.; Neuls, A.S.; Warner, C.L.

    1976-01-01

    A controlled-air incinerator and an associated high-energy aqueous off-gas cleaning system are being installed at the Los Alamos Scientific Laboratory (LASL) Transuranic Waste Treatment Development Facility (TDF) for evaluation as a low-level transuranic-contaminated (TRU) solid waste volume reduction process. Program objectives are: (1) assembly and operation of a production scale (45 kg/hr) operation of ''off-the-shelf'' components representative of current incineration and pollution control technology; (2) process development and modification to meet radioactive health and safety standards, and (3) evaluation of the process to define the advantages and limitations of conventional technology. The results of the program will be the design specifications and operating procedures necessary for successful incineration of TRU waste. Testing, with nonradioactive waste, will begin in October 1976. This discussion covers commercially available incinerator and off-gas cleaning components, the modifications required for radioactive service, process components performance expectations, and a description of the LASL experimental program

  3. Development of thermal conditioning technology for Alpha-containment wastes: Alpha-contaminated waste incineration technology

    International Nuclear Information System (INIS)

    Kim, Joon Hyung; Kim, Jeong Guk; Yang, Hee Chul; Choi, Byung Seon; Jeong, Myeong Soo

    1999-03-01

    As the first step of a 3-year project named 'development of alpha-contaminated waste incineration technology', the basic information and data were reviewed, while focusing on establishment of R and D direction to develop the final goal, self-supporting treatment of α- wastes that would be generated from domestic nuclear industries. The status on α waste incineration technology of advanced states was reviewed. A conceptual design for α waste incineration process was suggested. Besides, removal characteristics of volatile metals and radionuclides in a low-temperature dry off-gas system were investigated. Radiation dose assessments and some modification for the Demonstration-scale Incineration Plant (DSIP) at Korea Atomic Energy Research Institute (KAERI) were also done

  4. Development of thermal conditioning technology for Alpha-containment wastes: Alpha-contaminated waste incineration technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Joon Hyung; Kim, Jeong Guk; Yang, Hee Chul; Choi, Byung Seon; Jeong, Myeong Soo

    1999-03-01

    As the first step of a 3-year project named 'development of alpha-contaminated waste incineration technology', the basic information and data were reviewed, while focusing on establishment of R and D direction to develop the final goal, self-supporting treatment of {alpha}- wastes that would be generated from domestic nuclear industries. The status on {alpha} waste incineration technology of advanced states was reviewed. A conceptual design for {alpha} waste incineration process was suggested. Besides, removal characteristics of volatile metals and radionuclides in a low-temperature dry off-gas system were investigated. Radiation dose assessments and some modification for the Demonstration-scale Incineration Plant (DSIP) at Korea Atomic Energy Research Institute (KAERI) were also done.

  5. Assessment of a spent fuel disposal canister. Assessment studies for a copper canister with cast steel inner component

    International Nuclear Information System (INIS)

    Bond, A.E.; Hoch, A.R.; Jones, G.D.; Tomczyk, A.J.; Wiggin, R.M.; Worraker, W.J.

    1997-05-01

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden, is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in vertical storage holes drilled in a series of caverns excavated from the granite bedrock at a depth of about 500 m. Each canister will be surrounded by compacted bentonite clay. In this report, a simple model of the behaviour of the canister subsequent to a first breach in its copper overpack is developed. This model is used to predict: -the ingress of water to the canister (as a function of the size and the shape of the initial defect, the buffer conductivity, the corrosion rate and the pressure inside the canister); -the build-up of corrosion products in the canister (as a function of the available water in the canister, the corrosion rate and the properties of the corrosion products); -the effect of corrosion on the structural integrity of the canister. A number of different scenarios for the location of the breach in the copper overpack are considered

  6. Commercial Nuclear Waste Research and Development Program. Annual report, fiscal year 1984

    International Nuclear Information System (INIS)

    1985-01-01

    This document is a report of fiscal year 1984 activities performed to meet task objectives of the Nevada Nuclear Waste Storage Investigations (NNWSI) Project Plan. Noteworthy accomplishments were: operation of a fuel temperature test which evaluates fuel performance in an environment expected for metal cask storage. This test is being performed for the DOE Richland Operations Dry Storage Fuel Integrity Demonstration Program; characterization of fuel assemblies and storage canisters, which includes fuel assembly visual inspection, videotape and still photography, and contamination swipe sampling; and canister gas and full volume filtration sampling, contamination swipe sampling, and residue collection; removal of fuel assemblies from seal welded storage canisters, using the E-MAD canister cutter; integrity monitoring of fuel assemblies in seal welded canisters, by sampling and analysis of canister cover gas; and fuel assemblies in unwelded canisters, by collection and analysis of fuel surface contamination swipe samples and canister residue; emplacement/removal of canisterized fuel assemblies in the E-MAD drywells; and preparation of a summary storage history of all fuel assemblies at the E-MAD facility, including FY 1983 activities involving those fuel assemblies

  7. Technologies for in situ immobilization and isolation of radioactive wastes at disposal and contaminated sites

    International Nuclear Information System (INIS)

    1997-11-01

    This report describes technologies that have been developed worldwide and the experiences applied to both waste disposal and contaminated sites. The term immobilization covers both solidification and embedding of wastes

  8. Corrosion resistance of copper canister weld material

    International Nuclear Information System (INIS)

    Gubner, Rolf; Andersson, Urban

    2007-03-01

    The proposed design for a final repository for spent fuel and other long-lived residues is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast iron insert fitted inside a copper canister. SKB has since several years developed manufacturing processes for the canister components using a network of manufacturers. For the encapsulation process SKB has built the Canister Laboratory to demonstrate and develop the encapsulation technique in full scale. The critical part of the encapsulation of spent fuel is the sealing of the canister which is done by welding the copper lid to the cylindrical part of the canister. Two welding techniques have been developed in parallel, Electron Beam Welding (EBW) and Friction Stir Welding (FSW). During the past two decades, SKB has developed the technology EBW at The Welding Institute (TWI) in Cambridge, UK. The development work at the Canister Laboratory began in 1999. In electron beam welding, a gun is used to generate the electron beam which is aimed at the joint. The beam heats up the material to the melting point allowing a fusion weld to be formed. The gun was developed by TWI and has a unique design for use at reduced pressure. The system has gone through a number of improvements under the last couple of years including implementation of a beam oscillation system. However, during fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities at the new grain boundaries that might pose adverse effects on the corrosion resistance of welds. As a new method for joining, SKB has been developing friction stir welding (FSW) for sealing copper canisters for spent nuclear fuel in cooperation with TWI since 1997. FSW was invented in 1991 at TWI and is a thermo

  9. Corrosion resistance of copper canister weld material

    Energy Technology Data Exchange (ETDEWEB)

    Gubner, Rolf; Andersson, Urban [Corrosion and Metals Research Institute, Sto ckholm (Sweden)

    2007-03-15

    The proposed design for a final repository for spent fuel and other long-lived residues is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast iron insert fitted inside a copper canister. SKB has since several years developed manufacturing processes for the canister components using a network of manufacturers. For the encapsulation process SKB has built the Canister Laboratory to demonstrate and develop the encapsulation technique in full scale. The critical part of the encapsulation of spent fuel is the sealing of the canister which is done by welding the copper lid to the cylindrical part of the canister. Two welding techniques have been developed in parallel, Electron Beam Welding (EBW) and Friction Stir Welding (FSW). During the past two decades, SKB has developed the technology EBW at The Welding Institute (TWI) in Cambridge, UK. The development work at the Canister Laboratory began in 1999. In electron beam welding, a gun is used to generate the electron beam which is aimed at the joint. The beam heats up the material to the melting point allowing a fusion weld to be formed. The gun was developed by TWI and has a unique design for use at reduced pressure. The system has gone through a number of improvements under the last couple of years including implementation of a beam oscillation system. However, during fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities at the new grain boundaries that might pose adverse effects on the corrosion resistance of welds. As a new method for joining, SKB has been developing friction stir welding (FSW) for sealing copper canisters for spent nuclear fuel in cooperation with TWI since 1997. FSW was invented in 1991 at TWI and is a thermo

  10. Rapid monitoring for transuranic contaminants during buried waste retrieval

    International Nuclear Information System (INIS)

    McIsaac, C.V.; Sill, C.W.; Gehrke, R.J.; Shaw, P.G.; Randolph, P.D.; Amaro, C.R.; Pawelko, R.J.; Thompson, D.N.; Loomis, G.G.

    1991-03-01

    This document reports results of research performed in support of possible future transuranic waste retrieval operations at the Idaho National Engineering Laboratory Radioactive Waste Management Complex. The focus of this research was to evaluate various methods of performing rapid and, as much as possible, ''on-line'' quantitative measurements of 239 Pu or 241 Am, either as airborne or loose contamination. Four different alpha continuous air monitors were evaluated for lower levels of detection of airborne 239 Pu. All of the continuous air monitors were evaluated by sampling ambient air. In addition, three of the continuous air monitors were evaluated by sampling air synthetically laden with clean dust and dust spiked with 239 Pu. Six methods for making quantitative measurements of loose contamination were investigated. They were: (1) microwave digestion followed by counting in a photon electron rejecting alpha liquid scintillation spectrometer, (2) rapid radiochemical separation followed by alpha spectrometry, (3) measurement of the 241 Am 59 keV gamma ray using a thin window germanium detector, (4) measurement of uranium L-shell x-rays, (5) gross alpha counting using a large-area Ag activated ZnS scintillator, and (6) direct counting of alpha particles using a large-area ionization chamber. 40 refs., 42 figs., 24 tabs

  11. Decontamination method for radiation-contaminated metal waste

    International Nuclear Information System (INIS)

    Suwa, Takeshi; Kuribayashi, Nobuhide; Yasumune, Taketoshi.

    1991-01-01

    In immersing radiation-contaminated metal wastes into a sulfuric acid solution thereby peeling and removing radioactive deposition cruds and dissolving the surface of the matrix metals to eliminate radioactive contaminants, when the potential of the sulfuric acid solution is shifted to a higher direction by more than a certain level due to the increase of the amount of metal ions leached from the cruds and the matrix material, the leached metal ions are electrolytically reduced to control the potential of the sulfuric acid solution to less than a predetermined potential level. Although the dissolving rate is increased as the concentration of the sulfuric acid solution is higher, it is preferably from 0.5 to 2 mol/l, since higher concentration increases the load on the waste liquid processing. Further, the temperature for solution is set to higher than a room temperature and, preferably from 50 to 90degC. Further, the potential level of the solution, although varies somewhat depending on the concentration of the leached metal ions and the temperature, is preferably controlled to less than 0.1 to 0.2 V. This can attain high decontaminating effect in a short period of time by using a sulfuric acid solution alone. (T.M.)

  12. Feasibility study for a DOE research and production fuel multipurpose canister

    International Nuclear Information System (INIS)

    Lopez, D.A.; Abbott, D.G.

    1994-02-01

    This is a report of the feasibility of multipurpose canisters for transporting, storing, and sing of Department of Energy research and production spent nuclear fuel. Six representative Department of Energy fuel assemblies were selected, and preconceptual canister designs were developed to accommodate these assemblies. The study considered physical interface, structural adequacy, criticality safety, shielding capability, thermal performance of the canisters, and fuel storage site infrastructure. The external envelope of the canisters was designed to fit within the overpack casks for commercial canisters being developed for the Department of Energy Office of Civilian Radioactive Waste Management. The budgetary cost of canisters to handle all fuel considered is estimated at $170.8M. One large conceptual boiling water reactor canister design, developed for the Office of Civilian Radioactive Waste Management, and two new canister designs can accommodate at least 85% of the volume of the Department of Energy fuel considered. Canister use minimizes public radiation exposure and is cost effective compared with bare fuel handling. Results suggest the need for additional study of issues affecting canister use and for conceptual design development of the three canisters

  13. Analysis of contamination conditions of the Joyo Waste Treatment Facility

    International Nuclear Information System (INIS)

    Yoshizawa, S.; Ishijima, N.; Tanimoto, K.

    1999-08-01

    Decontamination methods have been studied for decommissioning of Joyo Waste Treatment Facility whose operation has been stopped in 1994. In this study, we analyzed samples of its system piping, whose dose rate was relatively low, to determine conditions of contamination. We also study appropriate decontamination methods for them. Results are as follows. 1. The inner surfaces of piping were covered with a very thin clad that was less than 1 micrometer in thickness and had many vacancies, looked like particle detachment, which were about 20 micrometers in depth. Something like corrosion product was observed near the surface and it was 440 micrometers in depth. 2. Radioactive contamination was considered to settle on a lower part of the piping and to be buried in the clad. A kind of dominant contamination nuclide was 60 Co. 3. Hot nitric acid process will be suitable for system decontamination to reduce dose rate before dismantling. But its feasibility tests are indispensable using samples of main system components that have high dose rate. Rubber lining tanks requires another methods because of its difficulty of decontamination. 4. Analyses and decontamination tests using main system are required to decide through decontamination methods according to the clearance level. (author)

  14. Processing method for contaminated water containing radioactive waste

    International Nuclear Information System (INIS)

    Tahara, Toshiaki; Fukagawa, Ken-ichiro.

    1994-01-01

    For absorbing contaminated water containing radioactive substances, a sheet is prepared by covering water absorbing pulps carrying an organic water absorbent having an excellent water absorbability is semi-solidified upon absorption water with a water permeable cloth, such as a non-woven fabric having a shape stability. As the organic water absorbent, a hydrophilic polymer which retains adsorbed water as it is used. In particular, a starch-grafted copolymer having an excellent water absorbability also for reactor water containing boric acid is preferred. The organic water absorbent can be carried on the water absorbing pulps by scattering a granular organic water absorbent to the entire surface of the water absorbing cotton pulp extended thinly to carry it uniformly and putting them between thin absorbing paper sheets. If contaminated water containing radioactive materials are wiped off by using such a sheet, the entire sheet is semi-solidified along with the absorption with no leaching of the contaminated water, thereby enabling to move the wastes to a furnace for applying combustion treatment. (T.M.)

  15. Gas generation from radiolytic attack of TRU-contaminated hydrogenous waste

    International Nuclear Information System (INIS)

    Zerwekh, A.

    1979-06-01

    In 1970, the Waste Management and Transportation Division of the Atomic Energy Commission ordered a segregation of transuranic (TRU)-contaminated solid wastes. Those below a contamination level of 10 nCi/g could still be buried; those above had to be stored retrievably for 20 y. The possibility that alpha-radiolysis of hydrogenous materials might produce toxic, corrosive, and flammable gases in retrievably stored waste prompted an investigation of gas identities and generation rates in the laboratory and field. Typical waste mixtures were synthesized and contaminated for laboratory experiments, and drums of actual TRU-contaminated waste were instrumented for field testing. Several levels of contamination were studied, as well as pressure, temperature, and moisture effects. G (gas) values were determined for various waste matrices, and degradation products were examined

  16. Status report, canister fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Claes-Goeran; Eriksson, Peter; Westman, Marika [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Emilsson, Goeran [CSM Materialteknik AB, Linkoeping (Sweden)

    2004-06-01

    The report gives an account of the development of material and fabrication technology for copper canisters with cast inserts during the period from 2000 until the start of 2004. The engineering design of the canister and the choice of materials in the constituent components described in previous status reports have not been significantly changed. In the reference canister, the thickness of the copper shell is 50 mm. Fabrication of individual components with a thinner copper thickness is done for the purpose of gaining experience and evaluating fabrication and inspection methods for such canisters. As a part of the development of cast inserts, computer simulations of the casting processes and techniques used at the foundries have been performed for the purpose of optimizing the material properties. These properties have been evaluated by extensive tensile testing and metallographic inspection of test material taken from discs cut at different points along the length of the inserts. The testing results exhibit a relatively large spread. Low elongation values in certain tensile test specimens are due to the presence of poorly formed graphite, porosities, slag or other casting defects. It is concluded in the report that it will not be possible to avoid some presence of observed defects in castings of this size. In the deep repository, the inserts will be exposed to compressive loading and the observed defects are not critical for strength. An analysis of the strength of the inserts and formulation of relevant material requirements must be based on a statistical approach with probabilistic calculations. This work has been initiated and will be concluded during 2004. An initial verifying compression test of a canister in an isostatic press has indicated considerable overstrength in the structure. Seamless copper tubes are fabricated by means of three methods: extrusion, pierce and draw processing, and forging. It can be concluded that extrusion tests have revealed a

  17. Status report, canister fabrication

    International Nuclear Information System (INIS)

    Andersson, Claes-Goeran; Eriksson, Peter; Westman, Marika; Emilsson, Goeran

    2004-06-01

    The report gives an account of the development of material and fabrication technology for copper canisters with cast inserts during the period from 2000 until the start of 2004. The engineering design of the canister and the choice of materials in the constituent components described in previous status reports have not been significantly changed. In the reference canister, the thickness of the copper shell is 50 mm. Fabrication of individual components with a thinner copper thickness is done for the purpose of gaining experience and evaluating fabrication and inspection methods for such canisters. As a part of the development of cast inserts, computer simulations of the casting processes and techniques used at the foundries have been performed for the purpose of optimizing the material properties. These properties have been evaluated by extensive tensile testing and metallographic inspection of test material taken from discs cut at different points along the length of the inserts. The testing results exhibit a relatively large spread. Low elongation values in certain tensile test specimens are due to the presence of poorly formed graphite, porosities, slag or other casting defects. It is concluded in the report that it will not be possible to avoid some presence of observed defects in castings of this size. In the deep repository, the inserts will be exposed to compressive loading and the observed defects are not critical for strength. An analysis of the strength of the inserts and formulation of relevant material requirements must be based on a statistical approach with probabilistic calculations. This work has been initiated and will be concluded during 2004. An initial verifying compression test of a canister in an isostatic press has indicated considerable overstrength in the structure. Seamless copper tubes are fabricated by means of three methods: extrusion, pierce and draw processing, and forging. It can be concluded that extrusion tests have revealed a

  18. Decontamination of alpha contaminated metallic waste by cerium IV redox process

    International Nuclear Information System (INIS)

    Shah, J.G.; Dhami, P.S.; Gandhi, P.M.; Wattal, P.K.

    2012-01-01

    Decontamination of alpha contaminated metallic waste is an important aspect in the management of waste generated during dismantling and decommissioning of nuclear facilities. Present work on cerium redox process targets decontamination of alpha contaminated metallic waste till it qualifies for the non alpha waste category for disposal in near surface disposal facility. Recovery of the alpha radio nuclides and cerium from aqueous secondary waste streams was also studied deploying solvent extraction process and established. The alpha-lean secondary waste stream has been immobilised in cement based matrix for final disposal. (author)

  19. Waste Tank Vapor Project: Vapor characterization of Tank 241-C-103: Report for SUMMA trademark canister samples received 11/29/93 (sample jobs 4 and 5)

    International Nuclear Information System (INIS)

    Clauss, T.R.; Lucke, R.B.; McVeety, B.; Allwine, K.J.; Fruchter, J.S.

    1994-09-01

    The purpose of Sample Jobs 4 and 5 was to determine whether the organic nitrites observed on the outside of tank 241-C-103 originated in the tank or from degradation products of the high-efficiency particulate air (HEPA) filter. The plan was to take samples from either side of the HE-PA filter. The relative level of organic nitrites would help determine whether they were produced in the filter or the tank. Pacific Northwest Laboratory was responsible for analyzing the SUMMA trademark canisters collected in support of this study. The laboratory was to analyze the SUMMA trademark Canister samples according to letters of instruction and report all semivolatile and volatile organic constituents detected in the tank headspace. Pacific Northwest Laboratory was also to submit a letter report to the Program Manager of all qualitative and quantitative analytical data, and estimate concentrations of any aliphatic nitrites identified. This was one of the first sampling activities for this program, and a number of errors were made both in the field and in the laboratory. Because of these errors, the samples and results were of questionable value. Therefore, Westinghouse program management asked that the analysis of the samples for this report not be completed. This report describes the few results that were generated before we were asked to stop work on this activity. In addition to analyzing SUMMA trademark canisters, PNL operates a site portable weather station near tank 241-C-103. Pacific Northwest Laboratory was required to collect atmospheric data starting 11/15/93, but the weather station was already collecting data during the time of both these two sample jobs (11/12/93 and 11/16/93). Therefore, a summary of the atmospheric data is also presented in this report

  20. K West Basin canister survey

    International Nuclear Information System (INIS)

    Pitner, A.L.

    1998-01-01

    A survey was conducted of the K West Basin to determine the distribution of canister types that contain the irradiated N Reactor fuel. An underwater camera was used to conduct the survey during June 1998, and the results were recorded on videotape. A full row-by-row survey of the entire basin was performed, with the distinction between aluminum and stainless steel Mark 1 canisters made by the presence or absence of steel rings on the canister trunions (aluminum canisters have the steel rings). The results of the survey are presented in tables and figures. Grid maps of the three bays show the canister lid ID number and the canister type in each location that contained fuel. The following abbreviations are used in the grid maps for canister type designation: IA = Mark 1 aluminum, IS = Mark 1 stainless steel, and 2 = Mark 2 stainless steel. An overall summary of the canister distribution survey is presented in Table 1. The total number of canisters found to contain fuel was 3842, with 20% being Mark 1 Al, 25% being Mark 1 SS, and 55% being Mark 2 SS. The aluminum canisters were predominantly located in the East and West bays of the basin

  1. Multi-Canister overpack internal HEPA filters

    International Nuclear Information System (INIS)

    SMITH, K.E.

    1998-01-01

    The rationale for locating a filter assembly inside each Multi-Canister Overpack (MCO) rather than include the filter in the Cold Vacuum Drying (CVD) process piping system was to eliminate the potential for contamination to the operators, processing equipment, and the MCO. The internal HEPA filters provide essential protection to facility workers from alpha contamination, both external skin contamination and potential internal depositions. Filters installed in the CVD process piping cannot mitigate potential contamination when breaking the process piping connections. Experience with K-Basin material has shown that even an extremely small release can result in personnel contamination and costly schedule disruptions to perform equipment and facility decontamination. Incorporating the filter function internal to the MCO rather than external is consistent with ALARA requirements of 10 CFR 835. Based on the above, the SNF Project position is to retain the internal HEPA filters in the MCO design

  2. Decontamination of uranium-contaminated waste oil using supercritical fluid and nitric acid

    International Nuclear Information System (INIS)

    Sung, J.; Kim, J.; Lee, Y.; Seol, J.; Ryu, J.; Park, K.

    2011-01-01

    The waste oil used in nuclear fuel processing is contaminated with uranium because of its contact with materials or environments containing uranium. Under current law, waste oil that has been contaminated with uranium is very difficult to dispose of at a radioactive waste disposal site. To dispose of the uranium-contaminated waste oil, the uranium was separated from the contaminated waste oil. Supercritical R-22 is an excellent solvent for extracting clean oil from uranium-contaminated waste oil. The critical temperature of R-22 is 96.15 deg. C and the critical pressure is 49.9 bar. In this study, a process to remove uranium from the uranium-contaminated waste oil using supercritical R-22 was developed. The waste oil has a small amount of additives containing N, S or P, such as amines, dithiocarbamates and dialkyldithiophosphates. It seems that these organic additives form uranium-combined compounds. For this reason, dissolution of uranium from the uranium-combined compounds using nitric acid was needed. The efficiency of the removal of uranium from the uranium-contaminated waste oil using supercritical R-22 extraction and nitric acid treatment was determined. (authors)

  3. Onsite disposal of radioactive waste: Estimating potential groundwater contamination

    International Nuclear Information System (INIS)

    Goode, D.J.; Neuder, S.M.; Pennifill, R.A.; Ginn, T.

    1986-11-01

    Volumes 1 and 2 of this report describe the NRC's methodology for assessing the potential public health and environmental impacts associated with onsite disposal of very low activity radioactive materials. This volume (Vol. 3) describes a general methodology for predicting potential groundwater contamination from onsite disposal. The methodology includes formulating a conceptual model, representing the conceptual model mathematically, estimating conservative parameters, and predicting receptor concentrations. Processes which must generally be considered in the methodology include infiltration, leaching of radionuclides from the waste, transport to the saturated zone, transport within the saturated zone, and withdrawal at a receptor location. A case study of shallow burial of iodine-125 illustrates application of the MOCMOD84 version of the US Geological Survey's 2-D solute transport model and a corresponding analytical solution. The appendices include a description and listing of MOCMOD84, descriptions of several analytical solution techniques, and a procedure for estimating conservative groundwater velocity values

  4. Microbial contamination level of air in animal waste utilization plants.

    Science.gov (United States)

    Chmielowiec-Korzeniowska, Anna; Tymczyna, Leszek; Drabik, Agata; Krzosek, Łukasz

    2016-01-01

    The aim of this research was evaluation of microbial contamination of air within and in the vicinity of animal waste disposal plants. Air samples were analyzed to determine total bacterial and fungal counts as well as microbial species composition. Measurements of climate conditions (temperature, humidity, air motion) and total dust concentration were also performed. Total numbers of bacteria and fungi surpassed the threshold limit values for production halls. The most abundant bacteria detected were those consisting of physiological microflora of animal dermis and mucosa. Fungal species composition proved to be most differentiated in the air beyond the plant area. Aspergillus versicolor, a pathogenic and allergenic filamentous fungus, was isolated only inside the rendering plant processing hall. The measurement results showed a low sanitary-hygienic state of air in the plant processing halls and substantial air pollution in its immediate vicinity.

  5. Waste material recycling: Assessment of contaminants limiting recycling

    DEFF Research Database (Denmark)

    Pivnenko, Kostyantyn

    systematically investigated. This PhD project provided detailed quantitative data following a consistent approach to assess potential limitations for the presence of chemicals in relation to material recycling. Paper and plastics were used as illustrative examples of materials with well-established recycling...... schemes and great potential for increase in recycling, respectively. The approach followed in the present work was developed and performed in four distinct steps. As step one, fractional composition of waste paper (30 fractions) and plastics (9 fractions) from households in Åbenrå municipality (Southern...... detrimental to their recycling. Finally, a material flow analysis (MFA) approach revealed the potential for accumulation and spreading of contaminants in material recycling, on the example of the European paper cycle. Assessment of potential mitigation measures indicated that prevention of chemical use...

  6. Hanford Site Tank 241-C-108 Residual Waste Contaminant Release Models and Supporting Data

    Energy Technology Data Exchange (ETDEWEB)

    Cantrell, Kirk J.; Krupka, Kenneth M.; Geiszler, Keith N.; Arey, Bruce W.; Schaef, Herbert T.

    2010-06-18

    This report presents the results of laboratory characterization, testing, and analysis for a composite sample (designated 20578) of residual waste collected from single-shell tank C-108 during the waste retrieval process after modified sluicing. These studies were completed to characterize concentration and form of contaminant of interest in the residual waste; assess the leachability of contaminants from the solids; and develop release models for contaminants of interest. Because modified sluicing did not achieve 99% removal of the waste, it is expected that additional retrieval processing will take place. As a result, the sample analyzed here is not expected to represent final retrieval sample.

  7. Canine scent detection and microbial source tracking of human waste contamination in storm drains.

    Science.gov (United States)

    Van De Werfhorst, Laurie C; Murray, Jill L S; Reynolds, Scott; Reynolds, Karen; Holden, Patricia A

    2014-06-01

    Human fecal contamination of surface waters and drains is difficult to diagnose. DNA-based and chemical analyses of water samples can be used to specifically quantify human waste contamination, but their expense precludes routine use. We evaluated canine scent tracking, using two dogs trained to respond to the scent of municipal wastewater, as a field approach for surveying human fecal contamination. Fecal indicator bacteria, as well as DNA-based and chemical markers of human waste, were analyzed in waters sampled from canine scent-evaluated sites (urban storm drains and creeks). In the field, the dogs responded positively (70% and 100%) at sites for which sampled waters were then confirmed as contaminated with human waste. When both dogs indicated a negative response, human waste markers were absent. Overall, canine scent tracking appears useful for prioritizing sampling sites for which DNA-based and similarly expensive assays can confirm and quantify human waste contamination.

  8. Effect of HNO3-cerium(IV) decontamination on stainless steel canister materials

    International Nuclear Information System (INIS)

    Westerman, R.E.; Mackey, D.B.

    1991-01-01

    Stainless steel canisters will be filled with vitrified radioactive waste at the West Valley Demonstration Project (WVDP), West Valley, NY. After they are filled, the sealed canisters will be decontaminated by immersion in a HNO 3 -Ce(IV) solution, which will remove the oxide film and a small amount of metal from the surface of the canisters. Studies were undertaken in support of waste form qualification activities to determine the effect of this decontamination treatment on the legibility of the weld-bead canister identification label, and to determine whether this decontamination treatment could induce stress-corrosion cracking (SCC) in the AISI 304L stainless steel (SS) canister material. Neither the label legibility nor the canister integrity with regard to SCC were found to be prejudiced by the simulated decontamination treatment

  9. Radiolytic gas generation in plutonium contaminated waste materials

    International Nuclear Information System (INIS)

    Kazanjian, A.R.

    1976-01-01

    Many plutonium contaminated waste materials decompose into gaseous products because of exposure to alpha radiation. The gases generated (usually hydrogen) over long-storage periods may create hazardous conditions. To determine the extent of such hazards, knowing the gas generation yields is necessary. These yields were measured by contacting some common Rocky Flats Plant waste materials with plutonium and monitoring the enclosed atmospheres for extensive periods of time. The materials were Plexiglas, polyvinyl chloride, glove-box gloves, machining oil, carbon tetrachloride, chlorothene VG solvent, Kimwipes (dry and wet), polyethylene, Dowex-1 resin, and surgeon's gloves. Both 239 Pu oxide and 238 Pu oxide were used as radiation sources. The gas analyses were made by mass spectrometry and the results obtained were the total gas generation, the hydrogen generation, the oxygen consumption rate, and the gas composition over the entire storage period. Hydrogen was the major gas produced in most of the materials. The total gas yields varied from 0.71 to 16 cm 3 (standard temperature pressure) per day per curie of plutonium. The oxygen consumption rates varied from 0.0088 to 0.070 millimoles per day per gram of plutonium oxide-239 and from 0.0014 to 0.0051 millimoles per day per milligram 238 Pu

  10. Incineration facility for radioactively contaminated polychlorinated biphenyls and other wastes

    International Nuclear Information System (INIS)

    1982-06-01

    The statement assesses the environmental impacts associated with the construction of an incineration facility and related support facilities for the disposal of hazardous organic waste materials (including PCBs) which are contaminated with trace quantities of low-assay enriched uranium. The proposed action includes the incineration facility at Oak Ridge, Tennessee and storage, packaging, and shipping facilities at the Gaseous Diffusion Plants in Paducah, KY, and Portsmouth, OH; hazardous organic wastes from these plants and from the Y-12 Plant and Oak Ridge National Laboratories would be shipped to the proposed incineration facility. Impacts assessed include the effects of the project on air and water quality, on socioeconomic conditions, on public and occupational health and safety, and on ecology. Additionally, the statement presents an assessment of the potential impacts from accidents at the incineration facility or during transportation of the waste materials to the facility. The major impact identified was the potential for short-term occupational exposure to high concentrations of PCBs in smoke during the worst credible accident; mitigation of this impact will be addressed during the final design of the proposed facility. Alternatives which were assessed include no action, chemical destruction processes, and alternative transportation routes; all would have greater adverse impact or would increase the risk of an accident with the potential for adverse impact. The alternatives of commercial disposal, alternative sites, multiple incinerators, and alternative modes were eliminated from detailed analysis either because they are not feasible or because preliminary analysis showed that they would have clearly more adverse impact upon the environment than the proposed action

  11. Modeling Organic Contaminant Desorption from Municipal Solid Waste Components

    Science.gov (United States)

    Knappe, D. R.; Wu, B.; Barlaz, M. A.

    2002-12-01

    Approximately 25% of the sites on the National Priority List (NPL) of Superfund are municipal landfills that accepted hazardous waste. Unlined landfills typically result in groundwater contamination, and priority pollutants such as alkylbenzenes are often present. To select cost-effective risk management alternatives, better information on factors controlling the fate of hydrophobic organic contaminants (HOCs) in landfills is required. The objectives of this study were (1) to investigate the effects of HOC aging time, anaerobic sorbent decomposition, and leachate composition on HOC desorption rates, and (2) to simulate HOC desorption rates from polymers and biopolymer composites with suitable diffusion models. Experiments were conducted with individual components of municipal solid waste (MSW) including polyvinyl chloride (PVC), high-density polyethylene (HDPE), newsprint, office paper, and model food and yard waste (rabbit food). Each of the biopolymer composites (office paper, newsprint, rabbit food) was tested in both fresh and anaerobically decomposed form. To determine the effects of aging on alkylbenzene desorption rates, batch desorption tests were performed after sorbents were exposed to toluene for 30 and 250 days in flame-sealed ampules. Desorption tests showed that alkylbenzene desorption rates varied greatly among MSW components (PVC slowest, fresh rabbit food and newsprint fastest). Furthermore, desorption rates decreased as aging time increased. A single-parameter polymer diffusion model successfully described PVC and HDPE desorption data, but it failed to simulate desorption rate data for biopolymer composites. For biopolymer composites, a three-parameter biphasic polymer diffusion model was employed, which successfully simulated both the initial rapid and the subsequent slow desorption of toluene. Toluene desorption rates from MSW mixtures were predicted for typical MSW compositions in the years 1960 and 1997. For the older MSW mixture, which had a

  12. Radioactive waste disposal package

    Science.gov (United States)

    Lampe, Robert F.

    1986-11-04

    A radioactive waste disposal package comprising a canister for containing vitrified radioactive waste material and a sealed outer shell encapsulating the canister. A solid block of filler material is supported in said shell and convertible into a liquid state for flow into the space between the canister and outer shell and subsequently hardened to form a solid, impervious layer occupying such space.

  13. Corrosion resistance of canisters for final disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Mattsson, E.

    1979-01-01

    A group of Swedish scientists has evaluated from the corrosion point of view three alternative canister types for final disposal of waste from nuclear reactors in boreholes in rock 500 m below ground. Titanium canisters with a wall-thickness of 6 mm and 100 mm thick lead lining have been estimated to have a life of at least thousands of years, and probably tens of thousands of years. Copper canisters with 200-mm-thick walls would last for hundreds of thousands of years. The third type, α-alumina sintered under isostatic pressure, is a very promising canister material

  14. Remote Welding, NDE and Repair of DOE Standardized Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Eric Larsen; Art Watkins; Timothy R. McJunkin; Dave Pace; Rodney Bitsoi

    2006-05-01

    The U.S. Department of Energy (DOE) created the National Spent Nuclear Fuel Program (NSNFP) to manage DOE’s spent nuclear fuel (SNF). One of the NSNFP’s tasks is to prepare spent nuclear fuel for storage, transportation, and disposal at the national repository. As part of this effort, the NSNFP developed a standardized canister for interim storage and transportation of SNF. These canisters will be built and sealed to American Society of Mechanical Engineers (ASME) Section III, Division 3 requirements. Packaging SNF usually is a three-step process: canister loading, closure welding, and closure weld verification. After loading SNF into the canisters, the canisters must be seal welded and the welds verified using a combination of visual, surface eddy current, and ultrasonic inspection or examination techniques. If unacceptable defects in the weld are detected, the defective sections of weld must be removed, re-welded, and re-inspected. Due to the high contamination and/or radiation fields involved with this process, all of these functions must be performed remotely in a hot cell. The prototype apparatus to perform these functions is a floor-mounted carousel that encircles the loaded canister; three stations perform the functions of welding, inspecting, and repairing the seal welds. A welding operator monitors and controls these functions remotely via a workstation located outside the hot cell. The discussion describes the hardware and software that have been developed and the results of testing that has been done to date.

  15. Research on corrosion aspects of the advanced cold process canister

    International Nuclear Information System (INIS)

    Blackwood, D.J.; Hoch, A.R.; Naish, C.C.; Rance, A.

    1994-01-01

    The Advanced Cold Process Canister (ACPC) is a waste canister being developed jointly by SKB and TVO for the disposal of spent nuclear fuel. It comprises an outer copper canister, with a carbon steel canister inside. A concern regarding the use of the ACPC is that, in the unlikely event that the outer copper canister is penetrated, the anaerobic corrosion of the carbon steel container may result in the formation of hydrogen gas bubbles. These bubbles could disrupt the backfill, and thus increase water flow through the near field and the flux of radionuclides to the host geology. A number of factors that influence the rate at which hydrogen evolves as a result of the anaerobic corrosion of carbon steel in artificial granitic groundwaters have been investigated. A previously observed, time-dependent decline in the hydrogen evolution rate has been confirmed as being due to the production of magnetite film. Once the magnetite film is about 0.7-1.0 μm thick, the rate of hydrogen evolution reaches a steady state value. The pH and the ionic strength of the groundwater were both found to influence the long-term hydrogen evolution rate. The results of the experimental programme were used to update a model of the corrosion behaviour and hydrogen production from the Advanced Cold Process Canister. 36 figs, 5 tabs, 13 refs

  16. Physical properties of encapsulate spent fuel in canisters

    International Nuclear Information System (INIS)

    1999-01-01

    Spent fuel and high-level wastes will be permanently stored in a deep geological repository (AGP). Prior to this, they will be encapsulated in canisters. The present report is dedicated to the study of such canisters under the different physical demands that they may undergo, be those in operating or accident conditions. The physical demands of interest include mechanical demands, both static and dynamic, and thermal demands. Consideration is given to the complete file of the canister, from the time when it is empty and without lid to the final conditions expected in the repository. Thermal analyses of canisters containing spent fuel are often carried out in two dimensions, some times with hypotheses of axial symmetry and some times using a plane transverse section through the centre of the canister. The results obtained in both types of analyses are compared here to those of complete three-dimensional analyses. The latter generate more reliable information about the temperatures that may be experienced by the canister and its contents; they also allow calibrating the errors embodied in the two-dimensional calculations. (Author)

  17. Grain boundary corrosion of copper canister weld material

    International Nuclear Information System (INIS)

    Gubner, Rolf; Andersson, Urban; Linder, Mats; Nazarov, Andrej; Taxen, Claes

    2006-01-01

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container fitted inside a corrosion-resistant copper canister. During fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow, they will tend to concentrate impurities within the copper at the new grain boundaries. The work described in this report was undertaken to determine whether there is any possibility of enhanced corrosion at grain boundaries within the copper canister, based on the recommendations of the report SKB-TR--01-09 (INIS ref. 32025363). Grain boundary corrosion of copper is not expected to be a problem for the copper canisters in a repository. However, as one step in the experimental verification it is necessary to study grain boundary corrosion of copper in an environment where it may occur. A literature study aimed to find one or several solutions that are aggressive with respect to grain boundary corrosion of copper. Copper specimens cut from welds of real copper canisters where exposed to aerated ammonium hydroxide solution for a period of 14 days at 80 degrees C and 10 bar pressure. The samples were investigated prior to exposure using the scanning Kelvin probe technique to characterize anodic and cathodic areas on the samples. The degree of corrosion was determined by optical microscopy. No grain boundary corrosion could be observed in the autoclave experiments, however, a higher rate of corrosion was observed for the weld material compared to the base material. The work suggests that grain boundary corrosion of copper weld material is most unlikely to adversely affect SKB's copper canisters under the conditions in the repository

  18. Grain boundary corrosion of copper canister weld material

    Energy Technology Data Exchange (ETDEWEB)

    Gubner, Rolf; Andersson, Urban; Linder, Mats; Nazarov, Andrej; Taxen, Claes [Corrosion and Metals Research Inst. (KIMAB), Stockholm (Sweden)

    2006-01-15

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container fitted inside a corrosion-resistant copper canister. During fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow, they will tend to concentrate impurities within the copper at the new grain boundaries. The work described in this report was undertaken to determine whether there is any possibility of enhanced corrosion at grain boundaries within the copper canister, based on the recommendations of the report SKB-TR--01-09 (INIS ref. 32025363). Grain boundary corrosion of copper is not expected to be a problem for the copper canisters in a repository. However, as one step in the experimental verification it is necessary to study grain boundary corrosion of copper in an environment where it may occur. A literature study aimed to find one or several solutions that are aggressive with respect to grain boundary corrosion of copper. Copper specimens cut from welds of real copper canisters where exposed to aerated ammonium hydroxide solution for a period of 14 days at 80 degrees C and 10 bar pressure. The samples were investigated prior to exposure using the scanning Kelvin probe technique to characterize anodic and cathodic areas on the samples. The degree of corrosion was determined by optical microscopy. No grain boundary corrosion could be observed in the autoclave experiments, however, a higher rate of corrosion was observed for the weld material compared to the base material. The work suggests that grain boundary corrosion of copper weld material is most unlikely to adversely affect SKB's copper canisters under the conditions in the repository.

  19. Status of the Multipurpose Canister (MPC) Project

    International Nuclear Information System (INIS)

    Hopper, J.P.

    1996-01-01

    The multipurpose canister (MPC) project represents a cornerstone of the current U.S. Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) program for handling spent nuclear fuel. The MPC and associated support equipment is being designed to accommodate the requirements for not only storage and transport but also for the specified disposal requirements of the mined geologic repository system. The phase 1 design effort for the MPC system, being performed by the Westinghouse team on behalf of TRW Environmental Safety Systems (TESS), the OCRWM management ampersand operating (M ampersand O) contractor, is on schedule for delivery of completed safety analysis reports (SARs) in April 1996

  20. National Enforcement Initiative: Preventing Animal Waste from Contaminating Surface and Ground Water

    Science.gov (United States)

    This page describes EPA's goal in preventing animal waste from contaminating surface and ground Water. It is an EPA National Enforcement Initiative. Both enforcement cases, and a map of enforcement actions are provided.

  1. Environmental health: an analysis of available and proposed remedies for victims of toxic waste contamination

    International Nuclear Information System (INIS)

    Hurwitz, W.J.

    1981-01-01

    Past and present residents of the Love Canal area near Niagara Falls, New York, fear that they and their homes have been contaminated by toxic wastes seeping out from nearby chemical disposal sites. Hundreds of landfills nationwide are as potentially dangerous as Love Canal. In the absence of a statutory remedy, victims of contamination must rely upon common law theories of lability in order to recover damages for injuries suffered as a result of toxic waste contamination. This Note examines the merits and deficiencies of four common law theories: negligence, strict liability, nuisance and trespass. The Note concludes that none of these remedies is adequate to assure recovery to a person injured by toxic waste disposal, and recommends that legislation be adopted to ensure that victims of toxic waste contamination can be compensated for their injuries

  2. Overview of advanced technologies for stabilization of 238Pu-contaminated waste

    International Nuclear Information System (INIS)

    Ramsey, K.B.; Foltyn, E.M.; Heslop, J.M.

    1998-02-01

    This paper presents an overview of potential technologies for stabilization of 238 Pu-contaminated waste. Los Alamos National Laboratory (LANL) has processed 238 PuO 2 fuel into heat sources for space and terrestrial uses for the past several decades. The 88-year half-life of 238 Pu and thermal power of approximately 0.6 watts/gram make this isotope ideal for missions requiring many years of dependable service in inaccessible locations. However, the same characteristic which makes 238 Pu attractive for heat source applications, the high Curie content (17 Ci/gram versus 0.06 Ci/gram for 239 Pu ), makes disposal of 238 Pu-contaminated waste difficult. Specifically, the thermal load limit on drums destined for transport to the Waste Isolation Pilot Plant (WIPP), 0.23 gram per drum for combustible waste, is impossible to meet for nearly all 238 Pu-contaminated glovebox waste. Use of advanced waste treatment technologies including Molten Salt Oxidation (MSO) and aqueous chemical separation will eliminate the combustible matrix from 238 Pu-contaminated waste and recover kilogram quantities of 238 PuO 2 from the waste stream. A conceptual design of these advanced waste treatment technologies will be presented

  3. Canisters and nonfuel components at commercial nuclear reactors

    International Nuclear Information System (INIS)

    Gibbard, K.; Disbrow, J.

    1994-01-01

    This paper discusses detailed data on canisters and nonfuel components (NFC) at US commercial nuclear power reactors. A wide variety of NFC have been reported on the Form RW-859, open-quotes Nuclear Fuel Dataclose quotes survey. They may have been integral with an assembly, noncanistered in baskets, destined for disposal as low-level radioactive waste, or stored in canisters. Similarly, data on the family of canistered spent nuclear fuel (SNF) in storage pools was compiled. Approximately 85 percent of the 40,194 pieces of nonfuel assembly (NFA) hardware reported were integral with an assembly. This represents data submitted by 95 of the 107 reactors in 10 generic assembly classes. In addition, a total of 286 canisters have been reported as being in storage pools as of December 31, 1992. However, an additional 264 open baskets were also reported to contain miscellaneous SNF and nonfuel materials, garbage and debris. All of these 286 canisters meet the dimensional envelope requirements specified for disposal for open-quotes standard fuelclose quotes under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (10 CFR 961); most of the baskets do not

  4. Aespoe Hard Rock Laboratory Canister Retrieval Test. Microorganisms in buffer from the Canister Retrieval Test - numbers and metabolic diversity

    Energy Technology Data Exchange (ETDEWEB)

    Lydmark, Sara; Pedersen, Karsten (Microbial Analytics Sweden AB (Sweden))

    2011-03-15

    'Canister Retrieval Test' (CRT) is an experiment that started at Aespoe Hard Rock Laboratory (HRL) 2000. CRT is a part of the investigations which evaluate a possible KBS-3 storage of nuclear waste. The primary aim was to see whether it is possible or not to retrieve a copper canister after storage under authentic KBS-3 conditions. However, CRT also provided a unique opportunity to investigate if bacteria survived in the bentonite buffer during storage. Therefore, in connection to the retrieval of the canister microbiological samples were extracted from the bentonite buffer and the bacterial composition was studied. In this report, microbiological analyses of a total of 66 samples at the C2, R10, R9 and R6 levels in the bentonite from CRT are presented and discussed. By culturing bacteria from the bentonite in specific media the following bacterial parameters were investigated: The total amount of culturable heterotrophic aerobic bacteria, sulphate-reducing bacteria, and bacteria that produce the organic compound acetate (acetogens). The biovolume in the bentonite was determined by detection of the ATP content. In addition, bacteria from the bentonite were cultured in different sulphate-reducing media. In these cultures, the presence of the biotic compounds sulphide and acetate was investigated, since these have potentially negative effect on the copper canister in a KBS-3 repository. The results were to some extent compared to density, water content, and temperature data provided by Clay Technology AB. The results showed that 100-102 viable sulphate-reducing and acetogenic bacteria and 102-104 heterotrophic aerobic bacteria g-1 bentonite were present after five years of storage in the rock. Bacteria with several morphologies could be found in the cultures with bentonite. The most bacteria were detected in the bentonite buffer close to the rock but in a few samples also in bentonite close to the copper canister. When the presence of bacteria in the

  5. Aespoe Hard Rock Laboratory Canister Retrieval Test. Microorganisms in buffer from the Canister Retrieval Test - numbers and metabolic diversity

    International Nuclear Information System (INIS)

    Lydmark, Sara; Pedersen, Karsten

    2011-03-01

    'Canister Retrieval Test' (CRT) is an experiment that started at Aespoe Hard Rock Laboratory (HRL) 2000. CRT is a part of the investigations which evaluate a possible KBS-3 storage of nuclear waste. The primary aim was to see whether it is possible or not to retrieve a copper canister after storage under authentic KBS-3 conditions. However, CRT also provided a unique opportunity to investigate if bacteria survived in the bentonite buffer during storage. Therefore, in connection to the retrieval of the canister microbiological samples were extracted from the bentonite buffer and the bacterial composition was studied. In this report, microbiological analyses of a total of 66 samples at the C2, R10, R9 and R6 levels in the bentonite from CRT are presented and discussed. By culturing bacteria from the bentonite in specific media the following bacterial parameters were investigated: The total amount of culturable heterotrophic aerobic bacteria, sulphate-reducing bacteria, and bacteria that produce the organic compound acetate (acetogens). The biovolume in the bentonite was determined by detection of the ATP content. In addition, bacteria from the bentonite were cultured in different sulphate-reducing media. In these cultures, the presence of the biotic compounds sulphide and acetate was investigated, since these have potentially negative effect on the copper canister in a KBS-3 repository. The results were to some extent compared to density, water content, and temperature data provided by Clay Technology AB. The results showed that 10 0 -10 2 viable sulphate-reducing and acetogenic bacteria and 10 2 -10 4 heterotrophic aerobic bacteria g -1 bentonite were present after five years of storage in the rock. Bacteria with several morphologies could be found in the cultures with bentonite. The most bacteria were detected in the bentonite buffer close to the rock but in a few samples also in bentonite close to the copper canister. When the presence of bacteria in the bentonite

  6. Incineration of technological waste contaminated with alpha emitters

    International Nuclear Information System (INIS)

    Otter, C.; Moncouyoux, J.P.; Cartier, R.; Durec, J.P.; Afettouche, R.

    1990-01-01

    A large R and D programme is in progress at the CEA on alpha-bearing waste incineration. The program is developed in the laboratory and a pilot plant including the following aspects: physico-chemical characterization of wastes, study of thermal decomposition of wastes, laboratory study of generated gases (first with inactive then with active wastes), development of an industrial pilot plant with inactive wastes, study of corrosion resistance of material (laboratory and pilot plant), study and qualification of nuclear measurements on wastes, ashes and equipment [fr

  7. Investigating cross-contamination by yeast strains from dental solid waste to waste-handling workers by DNA sequencing.

    Science.gov (United States)

    Vieira, Cristina Dutra; Tagliaferri, Thaysa Leite; de Carvalho, Maria Auxiliadora Roque; de Resende-Stoianoff, Maria Aparecida; Holanda, Rodrigo Assuncao; de Magalhães, Thais Furtado Ferreira; Magalhães, Paula Prazeres; Dos Santos, Simone Gonçalves; de Macêdo Farias, Luiz

    2018-04-01

    Trying to widen the discussion on the risks associated with dental waste, this study proposed to investigate and genetically compare yeast isolates recovered from dental solid waste and waste workers. Three samples were collected from workers' hands, nasal mucosa, and professional clothing (days 0, 30, and 180), and two from dental waste (days 0 and 180). Slide culture, microscopy, antifungal drug susceptibility, intersimple sequence repeat analysis, and amplification and sequencing of internal transcribed spacer regions were performed. Yeast strains were recovered from all waste workers' sites, including professional clothes, and from waste. Antifungal susceptibility testing demonstrated that some yeast recovered from employees and waste exhibited nonsusceptible profiles. The dendrogram demonstrated the presence of three major clusters based on similarity matrix and UPGMA grouping method. Two branches displayed 100% similarity: three strains of Candida guilliermondii isolated from different employees, working in opposite work shifts, and from diverse sites grouped in one part of branch 1 and cluster 3 that included two samples of Candida albicans recovered from waste and the hand of one waste worker. The results suggested the possibility of cross-contamination from dental waste to waste workers and reinforce the need of training programs focused on better waste management routines. © 2017 The Authors. MicrobiologyOpen published by John Wiley & Sons Ltd.

  8. OCRWM Bulletin: Westinghouse begins designing multi-purpose canister

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This publication consists of two parts: OCRWM (Office of Civilian Radioactive Waste Management) Bulletin; and Of Mountains & Science which has articles on the Yucca Mountain project. The OCRWM provides information about OCRWM activities and in this issue has articles on multi-purpose canister design, and transportation cask trailer.

  9. OCRWM Bulletin: Westinghouse begins designing multi-purpose canister

    International Nuclear Information System (INIS)

    1995-01-01

    This publication consists of two parts: OCRWM (Office of Civilian Radioactive Waste Management) Bulletin; and Of Mountains ampersand Science which has articles on the Yucca Mountain project. The OCRWM provides information about OCRWM activities and in this issue has articles on multi-purpose canister design, and transportation cask trailer

  10. ASSESSMENT OF RADIOACTIVE AND NON-RADIOACTIVE CONTAMINANTS FOUND IN LOW LEVEL RADIOACTIVE WASTE STREAMS

    International Nuclear Information System (INIS)

    R.H. Little, P.R. Maul, J.S.S. Penfoldag

    2003-01-01

    This paper describes and presents the findings from two studies undertaken for the European Commission to assess the long-term impact upon the environment and human health of non-radioactive contaminants found in various low level radioactive waste streams. The initial study investigated the application of safety assessment approaches developed for radioactive contaminants to the assessment of nonradioactive contaminants in low level radioactive waste. It demonstrated how disposal limits could be derived for a range of non-radioactive contaminants and generic disposal facilities. The follow-up study used the same approach but undertook more detailed, disposal system specific calculations, assessing the impacts of both the non-radioactive and radioactive contaminants. The calculations undertaken indicated that it is prudent to consider non-radioactive, as well as radioactive contaminants, when assessing the impacts of low level radioactive waste disposal. For some waste streams with relatively low concentrations of radionuclides, the potential post-closure disposal impacts from non-radioactive contaminants can be comparable with the potential radiological impacts. For such waste streams there is therefore an added incentive to explore options for recycling the materials involved wherever possible

  11. Prolixe-prototype reprocessing unit for irradiating wastes contamined with alpha emitters

    International Nuclear Information System (INIS)

    Madic, C.; Sontag, R.

    1987-01-01

    A large number of hot cells are employed for research on nuclear fuel reprocessing and the production of isotope of transuranium elements. These activities generate solid wastes highly contaminated with alpha, beta, gamma emitters. The Prolixe hot cell was built in order to: 1/ reprocess the solid wastes contaminated with alpha, beta, gamma emitters produced in the Radiochemistry building: 2/ produce package wastes storable in shallow-ground disposal sites: 3/ develop a process sufficiently flexible to make it applicable to waste produced in other installations. The process is based on waste leaching after grinding. Depending on the type of wastes the leaching reactant will have a different composition 1/ nitric acid solution for cellulose waste: 2/ nitric solutions containing Ag(II) for other material. The complete process should achieve: 1/ a high waste volume reduction factor: 2/ the production of immobilized waste packages storage in shallow-ground disposal sites: 3/ the recycling of transuranium elements: 4/ the generation of a minimal volume of effluents. This process can be considered as an alternative process to incineration for the reprocessing of solid wastes highly contaminated with alpha, beta, gamma emitters

  12. Treatment of Radioactive Contaminated Soil and Concrete Wastes Using the Regulatory Clearance

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Il Sik; Ryu, W. S.; Kim, T. K.; Shon, J. S.; Ahn, S. J.; Lee, Y. H.; Bae, S. M.; Hong, D. S.; Ji, Y. Y.; Lee, B. C

    2008-11-15

    In the radioactive waste storage facilities at the Korea Atomic Energy Research Institute (KAERI) in Daejoen, there are thousands drums of radioactive contaminated soil and concrete wastes. The soil and concrete wastes were generated in 1988 during the decommissioning process of the research reactor and the attached radioactive waste treatment facility which were located in Seoul. The wastes were transported to Daejeon and have been stored since then. At the generation time, the radioactive contamination of the wastes was very low, and the radionuclides in the wastes was Co-60 and Cs-137. As the wastes have been stored for more than 20 years, the radioactivity concentration of the wastes has been decayed to become very extremely low. The wastes are needed to be treated because they take up large spaces at the storage facility. Also by treating the wastes, final disposal cost can be saved. So, the regulatory clearance was considered as a treatment method for the soil and concrete wastes with extremely low radioactivity concentration.

  13. Overview of management programs for plutonium-contaminated solid waste in the U.S.A

    International Nuclear Information System (INIS)

    Ramsey, R.W. Jr.; Daly, G.H.

    1975-01-01

    Programs for transuranium-contaminated solid wastes (TRU) in the U.S.A. are emphasizing a reduction in waste generation and the development of appropriate treatments to reduce the volume of wastes requiring interim storage and final disposal. Research and Development is emphasizing the establishment of sufficient information on treatment, hazards and storage to adopt a standardized procedure for handling wastes during an interim retrievable period and for final disposal. Federal responsibility for TRU waste is being proposed except for minimum amounts acceptable for commercial burial

  14. Earthworms as colonisers: Primary colonisation of contaminated land, and sediment and soil waste deposits

    NARCIS (Netherlands)

    Eijsackers, H.J.P.

    2010-01-01

    This paper reviews the role of earthworms in the early colonisation of contaminated soils as well as sediment and waste deposits, which are worm-free because of anthropogenic activities such as open-cast mining, soil sterilisation, consistent pollution or remediation of contaminated soil. Earthworms

  15. MANAGING ARSENIC CONTAMINATED SOIL, SEDIMENT, AND INDUSTRIAL WASTE WITH SOLIDIFICATION/STABILIZATION TREATMENT

    Science.gov (United States)

    Arsenic contamination of soil, sediment and groundwater is a widespread problem in certain areas and has caused great public concern due to increased awareness of the health risks. Often the contamination is naturally occurring, but it can also be a result of waste generated from...

  16. CHEMICAL MARKERS OF HUMAN WASTE CONTAMINATION: ANALYSIS OF UROBILIN AND PHARMACEUTICALS IN SOURCE WATERS

    Science.gov (United States)

    Giving public water authorities another tool to monitor and measure levels of human waste contamination of waters simply and rapidly would enhance public protection. Most of the methods used today detect such contamination by quantifying microbes occurring in feces in high enough...

  17. Transport of multiassembly sealed canisters

    International Nuclear Information System (INIS)

    Quinn, R.D.; Lehnert, R.A.; Rosa, J.M.

    1992-01-01

    A significant portion of the commercial spent nuclear fuel in dry storage in the US will be stored in multiassembly sealed canisters before the DOE begins accepting fuel from utilities in 1998. This paper reports that it is desirable from economic and ALARA perspectives to transfer these canisters directly from the plant to the MRS. To this end, it is necessary that the multiassembly sealed canisters, which have been licensed for storage under 10CFR72, be qualified for shipment within a suitable shipping cask under the rules of 10CFR71. Preliminary work performed to date indicates that it is feasible to license a current canister design for transportation, and work is proceeding on obtaining NRC approval

  18. Demonstration of a Solution Film Leak Test Technique and Equipment for the S00645 Canister Closure

    International Nuclear Information System (INIS)

    Cannell, G.R.

    1999-01-01

    The purpose of this effort was to demonstrate that the SFT technique, when adapted to a DWPF canister nozzle, is capable of detecting leaks not meeting the Waste Acceptance Product Specifications (WAPS) acceptance criterion

  19. Assessment of the Characteristic Aggregates during a Decontamination of Contaminated Concrete Waste

    International Nuclear Information System (INIS)

    Min, B. Y.; Choi, W. K.; Oh, W. Z.; Jung, C. H.; Park, J. W.

    2008-01-01

    During a decommissioning of nuclear plants and facilities, large quantities of slightly contaminated concrete wastes are generated. The exposure to radiation over many years could be hazardous to human health. In Korea, the decontamination and decommissioning of the retired TRIGA MARK II and III research reactors and a uranium conversion plant at the Korea Atomic Energy Research Institute (KAERI) has been under way. Hundreds of tons of concrete wastes are expected from the D and D of these facilities. Typically, the contaminated layer is only 1∼10mm thick because cementitious materials are porous media, the penetration of radionuclides may occur up to several centimeters from the surface of a material. Contaminated concrete waste can be of two forms, either a surface or bulk contamination. Bulk contamination usually arises from a neutron activation of nuclides during the service life on a component. Surface activity can be a loose contamination arising from a deposition of nuclides from an interfacing medium, and it also can be tightly bound. Most of the dismantled concrete wastes are slightly contaminated rather than activated. This decontamination can be accomplished during the course of a separation of the concrete wastes contaminated with radioactive materials through a thermal treatment step of the radionuclide (e.g. cesium and strontium), transportation of the radionuclide to fine aggregates through a mechanical treatment step such as a crushing, milling and sieving. Produced fine powder (paste) should be stabilized for the final disposal. Melting technology has been known as the one of the most effective technologies for a stabilization and volume reduction to the paste. Therefore, a melting may be a last step in the decontamination of a contaminated paste. The aim of this study was to establish the separation conditions for an optimum decontamination for the treatment of concrete wastes contaminated with radionuclides. The separation tests had been

  20. Low-waste technology of prevention, decontamination and localization of radioactive contamination

    International Nuclear Information System (INIS)

    Kizhnerov, L. V.; Konstantinov, Ye. A.; Prokopenko, V. A.; Sorokin, N. M.

    1997-01-01

    The report presents the results of research in developing a low-waste technology of prevention, decontamination and localization of radioactive contamination founded on the of easily removed protective polymeric coating based on water and alcohol latexes and dispersion of polymers with special activating additives. The developed technology provides for the reduction of weakly fixed radioactive contamination of non-painted and painted surfaces to admissible levels (as a rule), it securely prevents and localizes contamination and does not generate secondary liquid radioactive wastes

  1. Corrosion resistance of metal materials for HLW canister

    International Nuclear Information System (INIS)

    Furuya, Takashi; Muraoka, Susumu; Tashiro, Shingo

    1982-02-01

    In order to verify the materials as an important artificial barrier for canister of vitrified high-level waste from spent fuel reprocessing, data and reports were researched on corrosion resistance of the materials under conditions from glass form production to final disposal. Then, in this report, investigated subjects, improvement methods and future subjects are reviewed. It has become clear that there would be no problem on the inside and outside corrosion of the canister during glass production, but long term corrosion and radiation effect tests and the vitrification methods would be subjects in future on interim storage and final disposal conditions. (author)

  2. Contamination by trace elements at e-waste recycling sites in Bangalore, India.

    Science.gov (United States)

    Ha, Nguyen Ngoc; Agusa, Tetsuro; Ramu, Karri; Tu, Nguyen Phuc Cam; Murata, Satoko; Bulbule, Keshav A; Parthasaraty, Peethmbaram; Takahashi, Shin; Subramanian, Annamalai; Tanabe, Shinsuke

    2009-06-01

    The recycling and disposal of electronic waste (e-waste) in developing countries is causing an increasing concern due to its effects on the environment and associated human health risks. To understand the contamination status, we measured trace elements (TEs) in soil, air dust, and human hair collected from e-waste recycling sites (a recycling facility and backyard recycling units) and the reference sites in Bangalore and Chennai in India. Concentrations of Cu, Zn, Ag, Cd, In, Sn, Sb, Hg, Pb, and Bi were higher in soil from e-waste recycling sites compared to reference sites. For Cu, Sb, Hg, and Pb in some soils from e-waste sites, the levels exceeded screening values proposed by US Environmental Protection Agency (EPA). Concentrations of Cr, Mn, Co, Cu, In, Sn, Sb, Tl, Pb and Bi in air from the e-waste recycling facility were relatively higher than the levels in Chennai city. High levels of Cu, Mo, Ag, Cd, In, Sb, Tl, and Pb were observed in hair of male workers from e-waste recycling sites. Our results suggest that e-waste recycling and its disposal may lead to the environmental and human contamination by some TEs. To our knowledge, this is the first study on TE contamination at e-waste recycling sites in Bangalore, India.

  3. Inventory of contaminants in waste wood; Inventering av foeroreningar i returtrae

    Energy Technology Data Exchange (ETDEWEB)

    Jermer, Joeran; Ekvall, Annika; Tullin, Claes [Swedish National Testing and Research Inst., Boraas (Sweden)

    2001-03-01

    Waste wood is increasingly used as fuel in Sweden. It is of Swedish origin as well as imported, mainly from Germany and the Netherlands. The waste wood is contaminated by e.g. paint and wood preservatives and objects of metal, glass, plastics etc. The contaminants may cause technical problems such as deposits and corrosion as well as plugging of air openings. The present study has focussed on potential contaminants in waste wood that could cause problems of technical as well as environmental nature. The major chemical contaminants are surface treatments (paints etc) and wood preservatives. The surface treatments contribute in particular to contaminants of zinc and lead. In some cases zinc has been found to cause severe deposits in the furnaces. Surface treatments also contribute to increased levels of sodium, chlorine, sulphur and nitrogen. Preservative-treated wood is the most important source of increased levels of copper, chromium and arsenic in the waste wood. Waste wood imported from Germany contains less arsenic but the same amount of copper and chromium as Swedish waste wood. The contents of mercury in German waste wood can be expected to be higher than in waste wood of Swedish origin. The fraction consisting of wood-based panels is comparably free from contaminants but as a result of the high contents of adhesives wood-based panels contribute to a higher proportion of nitrogen in waste wood than in forest residues. A great number of non-wood compounds (such as plastics and metals) do also contaminate waste wood. By careful and selective demolition and various sorting procedures most non-wood compounds will be separated from the waste wood. Waste sorting analyses carried out indicate that the waste wood contains approximately 1% non-wood compounds, mainly plastic and metal compounds, glass, dirt, concrete, bricks and gypsum. This may seem to be a small proportion, but if large amounts of waste wood are incinerated the non-wood compounds will inevitably cause

  4. Automated methodology for estimating waste streams generated from decommissioning contaminated facilities

    International Nuclear Information System (INIS)

    Toth, J.J.; King, D.A.; Humphreys, K.K.; Haffner, D.R.

    1994-01-01

    As part of the DOE Programmatic Environmental Impact Statement (PEIS), a viable way to determine aggregate waste volumes, cost, and direct labor hours for decommissioning and decontaminating facilities is required. In this paper, a methodology is provided for determining waste streams, cost and direct labor hours from remediation of contaminated facilities. The method is developed utilizing U.S. facility remediation data and information from several decommissioning programs, including reactor decommissioning projects. The method provides for rapid, consistent analysis for many facility types. Three remediation scenarios are considered for facility D ampersand D: unrestricted land use, semi-restricted land use, and restricted land use. Unrestricted land use involves removing radioactive components, decontaminating the building surfaces, and demolishing the remaining structure. Semi-restricted land use involves removing transuranic contamination and immobilizing the contamination on-site. Restricted land use involves removing the transuranic contamination and leaving the building standing. In both semi-restricted and restricted land use scenarios, verification of containment with environmental monitoring is required. To use the methodology, facilities are placed in a building category depending upon the level of contamination, construction design, and function of the building. Unit volume and unit area waste generation factors are used to calculate waste volumes and estimate the amount of waste generated in each of the following classifications: low-level, transuranic, and hazardous waste. Unit factors for cost and labor hours are also applied to the result to estimate D ampersand D cost and labor hours

  5. Mechanical Integrity of Canisters Using a Fracture Mechanics Approach

    Energy Technology Data Exchange (ETDEWEB)

    Koyama, Tomofumi; Guoxiang Zhang; Lanru Jing [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Land and Water Resources Engineering

    2006-07-15

    This report presents the methods and results of a research project about numerical modeling of mechanical integrity of cast-iron canisters for the final disposal of spent nuclear fuel in Sweden, using combined boundary element (BEM) and finite element (FEM) methods. The objectives of the project are: 1) to investigate the possibility of initiation and growth of fractures in the cast-iron canisters under the mechanical loading conditions defined in the premises of canister design by Swedish Nuclear Fuel and Waste Management Co. (SKB); 2) to investigate the maximum bearing capacity of the cast iron canisters under uniformly distributed and gradually increasing boundary pressure until plastic failure. Achievement of the two objectives may provide some quantitative evidence for the mechanical integrity and overall safety of the cast-iron canisters that are needed for the final safety assessment of the geological repository of the radioactive waste repository in Sweden. The geometrical dimension, distribution and magnitudes of loads and Material properties of the canisters and possible fractures were provided by the latest investigations of SKB. The results of the BEM simulations, using the commercial code BEASY, indicate that under the currently defined loading conditions the possibility of initiation of new fractures or growth of existing fractures (defects) are very small, due to the reasons that: 1) the canisters are under mainly compressive stresses; 2) the induced tensile stress regions are too small in both dimension and magnitude to create new fractures or to induce growth of existing fractures, besides the fact that the toughness of the fractures in the cast iron canisters are much higher that the stress intensity factors in the fracture tips. The results of the FEM simulation show a approximately 75 MPa maximum pressure beyond which plastic collapse of the cast-iron canisters may occur, using an elastoplastic Material model. This figure is smaller compared

  6. Transuranic contaminated waste form characterization and data base

    International Nuclear Information System (INIS)

    McArthur, W.C.; Kniazewycz, B.G.

    1980-07-01

    This report outlines the sources, quantities, characteristics and treatment of transuranic wastes in the United States. This document serves as part of the data base necessary to complete preparation and initiate implementation of transuranic wastes, waste forms, waste container and packaging standards and criteria suitable for inclusion in the present NRC waste management program. No attempt is made to evaluate or analyze the suitability of one technology over another. Indeed, by the nature of this report, there is little critical evaluation or analysis of technologies because such analysis is only appropriate when evaluating a particular application or transuranic waste streams. Due to fiscal restriction, the data base is developed from a myriad of technical sources and does not necessarily contain operating experience and the current status of all technologies. Such an effort was beyond the scope of this report

  7. Hanford site implementation plan for buried, transuranic-contaminated waste

    International Nuclear Information System (INIS)

    1987-05-01

    The GAO review of DOE's Defense Waste Management Plan (DWMP) identified deficiencies and provided recommendations. This report responds to the GAO recommendations with regard to the Hanford Site. Since the issuance of the DWMP, an extensive planning base has been developed for all high-level and transuranic waste at the Hanford Site. Thirty-three buried sites have been identified as possibly containing waste that can be classified as transuranic waste. Inventory reports and process flowsheets were used to provide an estimate of the radionuclide and hazardous chemical content of these sites and approximately 370 additional sites that can be classified as low-level waste. A program undertaken to characterize select sites suspected of having TRU waste to refine the inventory estimates. Further development and evaluation are ongoing to determine the appropriate remedial actions, with the objectives of balancing long-term risks with costs and complying with regulations. 18 refs., 7 figs., 6 tabs

  8. Transuranic contaminated waste form characterization and data base

    Energy Technology Data Exchange (ETDEWEB)

    McArthur, W.C.; Kniazewycz, B.G.

    1980-07-01

    This report outlines the sources, quantities, characteristics and treatment of transuranic wastes in the United States. This document serves as part of the data base necessary to complete preparation and initiate implementation of transuranic wastes, waste forms, waste container and packaging standards and criteria suitable for inclusion in the present NRC waste management program. No attempt is made to evaluate or analyze the suitability of one technology over another. Indeed, by the nature of this report, there is little critical evaluation or analysis of technologies because such analysis is only appropriate when evaluating a particular application or transuranic waste streams. Due to fiscal restriction, the data base is developed from a myriad of technical sources and does not necessarily contain operating experience and the current status of all technologies. Such an effort was beyond the scope of this report.

  9. COMSOL Multiphysics Model for HLW Canister Filling

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-11

    The U.S. Department of Energy (DOE) is building a Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is being temporarily stored in 177 underground tanks. Efforts are being made to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. Wastes containing high concentrations of Al2O3 and Na2O can contribute to nepheline (generally NaAlSiO4) crystallization, which can sharply reduce the chemical durability of high level waste (HLW) glass. Nepheline crystallization can occur during slow cooling of the glass within the stainless steel canister. The purpose of this work was to develop a model that can be used to predict temperatures of the glass in a WTP HLW canister during filling and cooling. The intent of the model is to support scoping work in the laboratory. It is not intended to provide precise predictions of temperature profiles, but rather to provide a simplified representation of glass cooling profiles within a full scale, WTP HLW canister under various glass pouring rates. These data will be used to support laboratory studies for an improved understanding of the mechanisms of nepheline crystallization. The model was created using COMSOL Multiphysics, a commercially available software. The model results were compared to available experimental data, TRR-PLT-080, and were found to yield sufficient results for the scoping nature of the study. The simulated temperatures were within 60 ºC for the centerline, 0.0762m (3 inch) from centerline, and 0.2286m (9 inch) from centerline thermocouples once the thermocouples were covered with glass. The temperature difference between the experimental and simulated values reduced to 40 ºC, 4 hours after the thermocouple was covered, and down to 20 ºC, 6 hours after the thermocouple was covered

  10. Proposal of a SiC disposal canister for very deep borehole disposal

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui-Joo; Lee, Minsoo; Lee, Jong-Youl; Kim, Kyungsu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this paper authors proposed a silicon carbide, SiC, disposal canister for the DBD concept in Korea. A. Kerber et al. first proposed the SiC canister for a geological disposal of HLW, CANDU or HTR spent nuclear fuels. SiC has some drawbacks in welding or manufacturing a large canister. Thus, we designed a double layered disposal canister consisting of a stainless steel outer layer and a SiC inner layer. KAERI has been interested in developing a very deep borehole disposal (DBD) of HLW generated from pyroprocessing of PWR spent nuclear fuel and supported the relevant R and D with very limited its own budget. KAERI team reviewed the DBD concept proposed by Sandia National Laboratories (SNL) and developed its own concept. The SNL concept was based on the steel disposal canister. The authors developed a new technology called cold spray coating method to manufacture a copper-cast iron disposal canister for a geological disposal of high level waste in Korea. With this method, 8 mm thin copper canister with 400 mm in diameter and 1200 mm in height was made. In general, they do not give any credit on the lifetime of a disposal canister in DBD concept unlike the geological disposal. In such case, the expensive copper canister should be replaced with another one. We designed a disposal canister using SiC for DBD. According to an experience in manufacturing a small size canister, the fabrication of a large-size one is a challenge. Also, welding of SiC canister is not easy. Several pathways are being paved to overcome it.

  11. Development of cold sprayed Cu coating for canister

    International Nuclear Information System (INIS)

    Kim, Hyung Jun; Kang, Yoon Ha

    2010-01-01

    Cold sprayed Cu deposition was studied for the application of outer part of canister for high level nuclear waste. Five commercially available pure Cu powders were analyzed and sprayed by high pressure cold spray system. Electrochemical corrosion test using potentiostat in 3.5% NaCl solution was conducted as well as microstructural analysis including hardness and oxygen content measurements. Overall evaluation of corrosion performance of cold sprayed Cu deposition is inferior to forged and extruded Cu plates, but some of Cu depositions are comparable to Cu plates. The simulated corrosion test in 200m underground cave is still in progress. The effect of cold spray process parameters was also studied and the results show that the type of nozzle is the most important other than powder feed rate, spray distance, and scan speed. 1/10 scale miniature of canister was manufactured confirming that the production of full scale canister is possible

  12. Test plan for the Sample Transfer Canister system

    International Nuclear Information System (INIS)

    Flanagan, B.D.

    1998-01-01

    The Sample Transfer Canister will be used by the Waste Receiving and Processing Facility (WRAP) for the transport of small quantity liquid samples that meet the definition of a limited quantity radioactive material, and may also be corrosive and/or flammable. These samples will be packaged and shipped in accordance with the US Department of Transportation (DOT) regulation 49 CFR 173.4, ''Exceptions for small quantities.'' The Sample Transfer Canister is of a ''French Can'' design, intended to be mated with a glove box for loading/unloading. Transport will typically take place north of the Wye Barricade between WRAP and the 222-S Laboratory. The Sample Transfer Canister will be shipped in an insulated ice chest, but the ice chest will not be a part of the small quantity package during prototype testing

  13. Contamination by perfluorinated compounds in water near waste recycling and disposal sites in Vietnam.

    Science.gov (United States)

    Kim, Joon-Woo; Tue, Nguyen Minh; Isobe, Tomohiko; Misaki, Kentaro; Takahashi, Shin; Viet, Pham Hung; Tanabe, Shinsuke

    2013-04-01

    There are very few reports on the contamination by perfluorinated chemicals (PFCs) in the environment of developing countries, especially regarding their emission from waste recycling and disposal sites. This is the first study on the occurrence of a wide range of PFCs (17 compounds) in ambient water in Vietnam, including samples collected from a municipal dumping site (MD), an e-waste recycling site (ER), a battery recycling site (BR) and a rural control site. The highest PFC concentration was found in a leachate sample from MD (360 ng/L). The PFC concentrations in ER and BR (mean, 57 and 16 ng/L, respectively) were also significantly higher than those detected in the rural control site (mean, 9.4 ng/L), suggesting that municipal solid waste and waste electrical and electronic equipment are potential contamination sources of PFCs in Vietnam. In general, the most abundant PFCs were perfluorooctanoic acid (PFOA), perfluorononanoic acid (PFNA), and perfluoroundecanoic acid (PFUDA; waste materials.

  14. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    International Nuclear Information System (INIS)

    Wishau, R.; Ramsey, K.B.; Montoya, A.

    1998-01-01

    This paper presents the technical and economic feasibility of molten salt oxidation technology as a volume reduction and recovery process for 238 Pu contaminated waste. Combustible low-level waste material contaminated with 238 Pu residue is destroyed by oxidation in a 900 C molten salt reaction vessel. The combustible waste is destroyed creating carbon dioxide and steam and a small amount of ash and insoluble 2328 Pu in the spent salt. The valuable 238 Pu is recycled using aqueous recovery techniques. Experimental test results for this technology indicate a plutonium recovery efficiency of 99%. Molten salt oxidation stabilizes the waste converting it to a non-combustible waste. Thus installation and use of molten salt oxidation technology will substantially reduce the volume of 238 Pu contaminated waste. Cost-effectiveness evaluations of molten salt oxidation indicate a significant cost savings when compared to the present plans to package, or re-package, certify and transport these wastes to the Waste Isolation Pilot Plant for permanent disposal. Clear and distinct cost advantages exist for MSO when the monetary value of the recovered 238 Pu is considered

  15. Mixed Waste Focus Area mercury contamination product line: An integrated approach to mercury waste treatment and disposal

    International Nuclear Information System (INIS)

    Hulet, G.A.; Conley, T.B.; Morris, M.I.

    1998-01-01

    The US Department of Energy (DOE) Mixed Waste Focus Area (MWFA) is tasked with ensuring that solutions are available for the mixed waste treatment problems of the DOE complex. During the MWFA's initial technical baseline development process, three of the top four technology deficiencies identified were related to the need for amalgamation, stabilization, and separation/removal technologies for the treatment of mercury and mercury-contaminated mixed waste. The focus area grouped mercury-waste-treatment activities into the mercury contamination product line under which development, demonstration, and deployment efforts are coordinated to provide tested technologies to meet the site needs. The Mercury Working Group (HgWG), a selected group of representatives from DOE sites with significant mercury waste inventories, is assisting the MWFA in soliciting, identifying, initiating, and managing efforts to address these areas. Based on the scope and magnitude of the mercury mixed waste problem, as defined by HgWG, solicitations and contract awards have been made to the private sector to demonstrate amalgamation and stabilization processes using actual mixed wastes. Development efforts are currently being funded under the product line that will address DOE's needs for separation/removal processes. This paper discusses the technology selection process, development activities, and the accomplishments of the MWFA to date through these various activities

  16. Preliminary Transportation, Aging and Disposal Canister System Performance Specification

    International Nuclear Information System (INIS)

    C.A Kouts

    2006-01-01

    This document provides specifications for selected system components of the Transportation, Aging and Disposal (TAD) canister-based system. A list of system specified components and ancillary components are included in Section 1.2. The TAD canister, in conjunction with specialized overpacks will accomplish a number of functions in the management and disposal of spent nuclear fuel. Some of these functions will be accomplished at purchaser sites where commercial spent nuclear fuel (CSNF) is stored, and some will be performed within the Office of Civilian Radioactive Waste Management (OCRWM) transportation and disposal system. This document contains only those requirements unique to applications within Department of Energy's (DOE's) system. DOE recognizes that TAD canisters may have to perform similar functions at purchaser sites. Requirements to meet reactor functions, such as on-site dry storage, handling, and loading for transportation, are expected to be similar to commercially available canister-based systems. This document is intended to be referenced in the license application for the Monitored Geologic Repository (MGR). As such, the requirements cited herein are needed for TAD system use in OCRWM's disposal system. This document contains specifications for the TAD canister, transportation overpack and aging overpack. The remaining components and equipment that are unique to the OCRWM system or for similar purchaser applications will be supplied by others

  17. Transuranic contaminated waste form characterization and data base

    International Nuclear Information System (INIS)

    Kniazewycz, B.G.; McArthur, W.C.

    1980-07-01

    This volume contains appendices A to F. The properties of transuranium (TRU) radionuclides are described. Immobilization of TRU wastes by bituminization, urea-formaldehyde polymers, and cements is discussed. Research programs at DOE facilities engaged in TRU waste characterization and management studies are described

  18. The main rules regarding the management of solid waste and liquid effluent contaminated during use at nuclear medicine departments

    International Nuclear Information System (INIS)

    Boudouin, E.

    2011-01-01

    This article describes the key requirements applicable to the management of contaminated medical waste and effluent from hospitals and health care centres, and more especially from nuclear medicine departments that use radionuclides for the purposes of diagnosis (in vivo or in vitro) or in patient treatment. It also presents the key management regulations, making a distinction between contaminated solid waste and contaminated liquid waste from such nuclear medicine departments. (author)

  19. Advances in encapsulation technologies for the management of mercury-contaminated hazardous wastes

    International Nuclear Information System (INIS)

    Randall, Paul; Chattopadhyay, Sandip

    2004-01-01

    Although industrial and commercial uses of mercury have been curtailed in recent times, there is a demonstrated need for the development of reliable hazardous waste management techniques because of historic operations that have led to significant contamination and ongoing hazardous waste generation. This study was performed to evaluate whether the U.S. EPA could propose treatment and disposal alternatives to the current land disposal restriction (LDR) treatment standards for mercury. The focus of this article is on the current state of encapsulation technologies that can be used to immobilize elemental mercury, mercury-contaminated debris, and other mercury-contaminated wastes, soils, sediments, or sludges. The range of encapsulation materials used in bench-scale, pilot-scale, and full-scale applications for mercury-contaminated wastes are summarized. Several studies have been completed regarding the application of sulfur polymer stabilization/solidification, chemically bonded phosphate ceramic encapsulation, and polyethylene encapsulation. Other materials reported in the literature as under development for encapsulation use include asphalt, polyester resins, synthetic elastomers, polysiloxane, sol-gels, Dolocrete TM , and carbon/cement mixtures. The primary objective of these encapsulation methods is to physically immobilize the wastes to prevent contact with leaching agents such as water. However, when used for mercury-contaminated wastes, several of these methods require a pretreatment or stabilization step to chemically fix mercury into a highly insoluble form prior to encapsulation. Performance data is summarized from the testing and evaluation of various encapsulated, mercury-contaminated wastes. Future technology development and research needs are also discussed

  20. The management and disposal of alpha-contaminated waste

    International Nuclear Information System (INIS)

    Duclos, J.; Farges, L.; Lavie, J.M.; Marque, Y.

    1981-01-01

    The establishment of the French National Agency for Radioactive Waste Management (ANDRA) in November 1979 marked the beginning of industrial management of this type of waste in France. The organization of this Agency is sufficiently flexible to reconcile the need for the assumption of responsibility by the public authorities for a matter having considerable long-term implications; the importance of making available to all radioactive-waste producers the benefits of the research carried out by large national entities; (Commissariat a l'energie atomique, Electricite de France, etc.) and the obligation to satisfy all the scientific and financial requirements regarding optimal radioactive-waste management. The Centre de stockage de la Manche (CSM) is at present concerned with the special requirements relating to alpha waste. These are being analysed, together with their implications for technical specifications and industrial management. A strategy for alpha waste storage is defined in the light of the forecasts of waste deliveries for the next 20 years. (author)

  1. Contaminant Release from Residual Waste in Single Shell Tanks at the Hanford Site, Washington, USA - 9276

    International Nuclear Information System (INIS)

    Cantrell, Kirk J.; Krupka, Kenneth M.; Deutsch, William J.; Lindberg, Michael J.

    2009-01-01

    Determinations of elemental and solid-phase compositions, and contaminant release studies have been applied in an ongoing study of residual tank wastes (i.e., waste remaining after final retrieval operations) from five of 149 underground single-shell storage tanks (241-C-103, 241-C-106, 241-C-202, 241-C-203, and 241-S-112) at the U.S. Department of Energy's Hanford Site in Washington State. This work is being conducted to support performance assessments that will be required to evaluate long-term health and safety risks associated with tank site closure. The results of studies completed to date show significant variability in the compositions, solid phase properties, and contaminant release characteristics from these residual tank wastes. This variability is the result of differences in waste chemistry/composition of wastes produced from several different spent fuel reprocessing schemes, subsequent waste reprocessing to remove certain target constituents, tank farm operations that concentrated wastes and mixed wastes between tanks, and differences in retrieval processes used to remove the wastes from the tanks. Release models were developed based upon results of chemical characterization of the bulk residual waste, solid-phase characterization (see companion paper 9277 by Krupka et al.), leaching and extraction experiments, and geochemical modeling. In most cases empirical release models were required to describe contaminant release from these wastes. Release of contaminants from residual waste was frequently found to be controlled by the solubility of phases that could not be identified and/or for which thermodynamic data and/or dissolution rates have not been measured. For example, significant fractions of Tc-99, I-129, and Cr appear to be coprecipitated at trace concentrations in metal oxide phases that could not be identified unambiguously. In the case of U release from tank 241-C-103 residual waste, geochemical calculations indicated that leachate

  2. Development of Decontamination Technology for Separating Radioactive Constituents from Contaminated Concrete Waste

    International Nuclear Information System (INIS)

    Min, B. Y.; Kim, G. N.; Lee, G. W.; Choi, W. K.; Jung, U. S.

    2010-01-01

    The large amount of contaminated concrete produced during decommissioning procedures and available decontamination. In Korea, more than more than 60 tons of concrete wastes contaminated with uranium compounds have been generated from UCP (Uranium Conversion Plant) by dismantling. A recycling or a volume reduction of the concrete wastes through the application of appropriate treatment technologies have merits from the view point of an increase in a resource recycling as well as a decrease in the amount of wastes to be disposed of resulting in a reduction of a disposal cost and an enhancement of the disposal safety. For unconditional release of building and reduction of radioactive concrete waste, mechanical methods and thermal stress methods have been selected. In the advanced countries, such as France, Japan, Germany, Sweden, and Belgium, techniques for reduction and reuse of the decommissioning concrete wastes have applied to minimize the total radioactive concrete waste volume by thermal and mechanical processes. It was found that volume reduction of contaminated concrete can be achieved by separation of the fine cement stone and coarse gravel. Typically, the contaminated layer is only 1∼10mm thick because cementitious materials are porous media, the penetration of radionuclides may occur up to several centimenters from the surface of a material. Most of the dismantled concrete wastes are slightly contaminated rather than activated. This decontamination can be accomplished during the course of a separation of the concrete wastes contaminated with radioactive materials through a thermal treatment step of the radionuclide (e.g. cesium and strontium), transportation of the radionuclide to fine aggregates through a mechanical treatment step. Concrete is a structural material which generally consists of a binder (cement), water, and aggregate. The interaction between highly charged calcium silicate hydrate (C-S-H) particles in the presence of divalent calcium

  3. Status of foreign practices for the management of alpha-contaminated radioactive wastes

    International Nuclear Information System (INIS)

    Lakey, L.T.

    1982-08-01

    Alpha-contaminated radioactive wastes, a product of mixed-oxide fuel fabrication, fuel reprocessing, weapons production and decommissioning programs, are being generated in at least ten countries. There is general agreement worldwide that these wastes should be treated differently than the beta-gamma or low-level waste. There is no consensus, however, on a quantitative definition of alpha-contaminated wastes. Reported definitions vary from > 0.035 nCi/g to > 100 nCi/g. Incineration is the most common treatment, with cement and bitumen the most common fixation agents. The only disposal means in use today are the sea dumping practice by Belgium and the United Kingdom and the surface disposal and deep-well discharge by the USSR. Sea dumping, however, is restricted to low levels of alpha activity, while the USSR appears to be favoring geologic disposal. All countries appear to be moving toward deep geologic repositories as the favored means of disposing of alpha-contaminated radioactive wastes. West Germany has actually disposed of such wastes in the Asse Salt Mine but has discontinued that operation for political reasons. Repository projects are actively under way in Belgium, West Germany, India, Sweden, and the Unted States, with many other countries planning repository programs. One US project, the Waste Isolation Pilot Plant, will, according to present schedules, be the first repository operational since Asse. 6 tables

  4. Review of information on the radiation chemistry of materials around waste canisters in salt and assessment of the need for additional experimental information

    Energy Technology Data Exchange (ETDEWEB)

    Jenks, G.H.; Baes, C.F. Jr.

    1980-03-01

    The brines, vapors, and salts precipitated from the brines will be exposed to gamma rays and to elevated temperatures in the regions close to a waste package in the salt. Accordingly, they will be subject to changes in composition brought about by reactions induced by the radiations and heat. This report reviews the status of information on the radiation chemistry of brines, gases, and solids which might be present around a waste package in salt and to assess the need for additional laboratory investigations on the radiation chemistry of these materials. The basic aspects of the radiation chemistry of water and aqueous solutions, including concentrated salt solutions, were reviewed briefly and found to be substantially unchanged from those presented in Jenks's 1972 review of radiolysis and hydrolysis in salt-mine brines. Some additional information pertaining to the radiolytic yields and reactions in brine solutions has become available since the previous review, and this information will be useful in the eventual, complete elucidation of the radiation chemistry of the salt-mine brines. 53 references.

  5. Shielding design of radioactive contaminated metal waste packaging

    International Nuclear Information System (INIS)

    Zou Wenhua; Dong Zhiqiang; Yao Zhenyu; Xu Shuhe; Wang Wen

    2015-01-01

    Focusing on the cylindrical source model to calculate γ dose field of waste packages with the relative formulae then derived. By comparing the calculated data of waste packages of type Ⅷ steel box with the monitoring data, it is found that the cylinder source model could accurately reflect the distributions of γ dose of the waste package. Based on the results of the cylindrical source model, a reasonable shielding technology applicable to waste package containers was designed to meet relevant requirements prescribed in standards about the transport and disposal of radioactive materials. The cylinder source model calculated dose distributions for single package in this paper is simple and easy to implement but slightly larger than the monitoring data providing a certain safety margin for the shielding design. It is suitable for radiological engineering practices. (authors)

  6. Degradation of organic contaminants found in organic waste

    DEFF Research Database (Denmark)

    Angelidaki, Irini; Mogensen, Anders Skibsted; Ahring, Birgitte Kiær

    2000-01-01

    In recent years, great interest has arisen in recycling of the waste created by modern society. A common way of recycling the organic fraction is amendment on farmland. However, these wastes may contain possible hazardous components in small amounts, which may prevent their use in farming. The ob...... phenol ethoxylates. The results are promising as they indicate that a great potential for biological degradation is present, though the inoculum containing the microorganisms capable of transforming the recalcitrant xenobiotics has to be chosen carefully....

  7. Radiological, physical, and chemical characterization of low-level alpha contaminated wastes stored at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Apel, M.L.; Becker, G.K.; Ragan, Z.K.; Frasure, J.; Raivo, B.D.; Gale, L.G.; Pace, D.P.

    1994-03-01

    This document provides radiological, physical, and chemical characterization data for low-level alpha-contaminated radioactive and low-level alpha-contaminated radioactive and hazardous (i.e., mixed) wastes stored at the Idaho National Engineering Laboratory and considered for treatment under the Private Sector Participation Initiative Program. Waste characterization data are provided in the form of INEL Waste Profile Sheets. These documents provide, for each content code, information on waste identification, waste description, waste storage configuration, physical/chemical waste composition, radionuclide and associated alpha activity waste characterization data, and hazardous constituents present in the waste. Information is provided for 97 waste streams which represent an estimated total volume of 25,450 m 3 corresponding to a total mass of approximately 12,000,000 kg. In addition, considerable information concerning alpha, beta, gamma, and neutron source term data specific to Rocky Flats-generated waste forms stored at the INEL are provided to assist in facility design specification

  8. Radiological, physical, and chemical characterization of low-level alpha contaminated wastes stored at the Idaho National Engineering Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Apel, M.L.; Becker, G.K.; Ragan, Z.K.; Frasure, J.; Raivo, B.D.; Gale, L.G.; Pace, D.P.

    1994-03-01

    This document provides radiological, physical, and chemical characterization data for low-level alpha-contaminated radioactive and low-level alpha-contaminated radioactive and hazardous (i.e., mixed) wastes stored at the Idaho National Engineering Laboratory and considered for treatment under the Private Sector Participation Initiative Program. Waste characterization data are provided in the form of INEL Waste Profile Sheets. These documents provide, for each content code, information on waste identification, waste description, waste storage configuration, physical/chemical waste composition, radionuclide and associated alpha activity waste characterization data, and hazardous constituents present in the waste. Information is provided for 97 waste streams which represent an estimated total volume of 25,450 m 3 corresponding to a total mass of approximately 12,000,000 kg. In addition, considerable information concerning alpha, beta, gamma, and neutron source term data specific to Rocky Flats-generated waste forms stored at the INEL are provided to assist in facility design specification.

  9. Evaluating non-incinerative treatment of organically contaminated low level mixed waste

    International Nuclear Information System (INIS)

    Shuck, D.L.; Wade, J.F.

    1993-01-01

    This investigation examines the feasibility of using non-incinerator technologies effectively to treat organically contaminated mixed waste. If such a system is feasible now, it would be easier to license because it would avoid the stigma that incineration has in the publics' perception. As other DOE facilities face similar problems, this evaluation is expected to be of interest to both DOE and the attendees of WM'93. This investigation considered treatment to land disposal restriction (LDR) standards of 21 different low level mixed (LLM) waste streams covered by the Rocky Flats Federal Facilities Compliance Agreement (FFCA) agreement with the Environmental Protection Agency (EPA). Typically the hazardous components consists of organic solvent wastes and the radioactive component consists of uranic/transuranic wastes. Limited amounts of cyanide and lead wastes are also involved. The primary objective of this investigation was to identify the minimum number of non-thermal unit processes needed to effectively treat this collection of mixed waste streams

  10. Reduction of radioactive waste from remediation of uranium-contaminated soil

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Il Gook; Kim, Seung Soo; Kim, Gye Nam; Han, Gyu Seong; Choi, Jong Won [Decontamination and Decommissioning Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-06-15

    Great amounts of solid radioactive waste (second waste) and waste solution are generated from the remediation of uranium-contaminated soil. To reduce these, we investigated washing with a less acidic solution and recycling the waste solution after removal of the dominant elements and uranium. Increasing the pH of the washing solution from 0.5 to 1.5 would be beneficial in terms of economics. A high content of calcium in the waste solution was precipitated by adding sulfuric acid. The second waste can be significantly reduced by using sorption and desorption techniques on ampholyte resin S-950 prior to the precipitation of uranium at pH 3.0.

  11. Reduction of radioactive waste from remediation of uranium-contaminated soil

    International Nuclear Information System (INIS)

    Kim, Il Gook; Kim, Seung Soo; Kim, Gye Nam; Han, Gyu Seong; Choi, Jong Won

    2016-01-01

    Great amounts of solid radioactive waste (second waste) and waste solution are generated from the remediation of uranium-contaminated soil. To reduce these, we investigated washing with a less acidic solution and recycling the waste solution after removal of the dominant elements and uranium. Increasing the pH of the washing solution from 0.5 to 1.5 would be beneficial in terms of economics. A high content of calcium in the waste solution was precipitated by adding sulfuric acid. The second waste can be significantly reduced by using sorption and desorption techniques on ampholyte resin S-950 prior to the precipitation of uranium at pH 3.0

  12. Mechanical Integrity of Copper Canister Lid and Cylinder. Sensitivity study

    International Nuclear Information System (INIS)

    Karlsson, Marianne

    2002-08-01

    This report is part of a study of the mechanical integrity of canisters used for disposal of nuclear fuel waste. The overall objective is to determine and ensure the static and long-term strength of the copper canister lid and cylinder casing. The canisters used for disposal nuclear fuel waste of type BWR consists of an inner part (insert) of ductile cast iron and an outer part of copper. The copper canister is to provide a sealed barrier between the contents of the canister and the surroundings. The study in this report complements the finite element analyses performed in an earlier study. The analyses aim to evaluate the sensitivity of the canister to tolerances regarding the gap between the copper cylinder and the cast iron insert. Since great uncertainties regarding the material's long term creep properties prevail, analyses are also performed to evaluate the effect of different creep data on the resulting strain and stress state. The report analyses the mechanical response of the lid and flange of the copper canister when subjected to loads caused by pressure from swelling bentonite and from groundwater at a depth of 500 meter. The loads acting on the canister are somewhat uncertain and the cases investigated in this report are possible cases. Load cases analysed are: Pressure 15 MPa uniformly distributed on lid and 5 MPa uniformly distributed on cylinder; Pressure 5 MPa uniformly distributed on lid and 15 MPa uniformly distributed on cylinder; Pressure 20 MPa uniformly distributed on lid and cylinder; and Side pressures 10 MPa and 20 MPa uniformly distributed on part of the cylinder. Creep analyses are performed for two of the load cases. For all considered designs high principal stresses appear on the outside of the copper cylinder in the region from the weld down to the level of the lid lower edge. Altering the gap between lid and cylinder and/or between cylinder and insert only marginally affects the resulting stress state. Fitting the lid in the cylinder

  13. Treatment of plutonium contamined solid wastes by electrogenerated Ag(II)

    International Nuclear Information System (INIS)

    Saulze, J.L.

    1990-01-01

    A process for the treatment of plutonium contaminated solid wastes is designed. Two types of wastes have been studied; incineration ashes (COGEMA UP1) and sludges produced in the cryotreatment facility in Cadarache Center (France). The principle of the process is based on the rapid dissolution of PuO 2 (contained in the wastes) under the action of aggressive Ag(II) species, regenerated electrochemically. In the case of the treatment of incinerator ashes an electrochemical pretreatment is necessary if the chloride ion content of the ashes is high. The feasibility of the decontamination process has been proved for the two types of plutonium contaminated solid wastes at a pilot level; for example 1 Kg of ashes (or 0.75 Kg of sludges) has been treated in one experiment, and 97% (or 95%) of the total plutonium was dissolved at the end of the experiment. Industrial applications of this new process are underway [fr

  14. Soil contamination by brominated flame retardants in open waste dumping sites in Asian developing countries.

    Science.gov (United States)

    Eguchi, Akifimi; Isobe, Tomohiko; Ramu, Karri; Tue, Nguyen Minh; Sudaryanto, Agus; Devanathan, Gnanasekaran; Viet, Pham Hung; Tana, Rouch Seang; Takahashi, Shin; Subramanian, Annamalai; Tanabe, Shinsuke

    2013-03-01

    In Asian developing countries, large amounts of municipal wastes are dumped into open dumping sites each day without adequate management. This practice may cause several adverse environmental consequences and increase health risks to local communities. These dumping sites are contaminated with many chemicals including brominated flame retardants (BFRs) such as polybrominated diphenyl ethers (PBDEs) and hexabromocyclododecanes (HBCDs). BFRs may be released into the environment through production processes and through the disposal of plastics and electronic wastes that contain them. The purpose of this study was to elucidate the status of BFR pollution in municipal waste dumping sites in Asian developing countries. Soil samples were collected from six open waste dumping sites and five reference sites in Cambodia, India, Indonesia, Malaysia, and Vietnam from 1999 to 2007. The results suggest that PBDEs are the dominant contaminants in the dumping sites in Asian developing countries, whereas HBCD contamination remains low. Concentrations of PBDEs and HBCDs ranged from ND to 180 μg/kg dry wt and ND to 1.4 μg/kg dry wt, respectively, in the reference sites and from 0.20 to 430 μg/kg dry wt and ND to 2.5 μg/kg dry wt, respectively, in the dumping sites. Contamination levels of PBDEs in Asian municipal dumping sites were comparable with those reported from electronic waste dismantling areas in Pearl River delta, China. Copyright © 2012 Elsevier Ltd. All rights reserved.

  15. Green remediation and recycling of contaminated sediment by waste-incorporated stabilization/solidification.

    Science.gov (United States)

    Wang, Lei; Tsang, Daniel C W; Poon, Chi-Sun

    2015-03-01

    Navigational/environmental dredging of contaminated sediment conventionally requires contained marine disposal and continuous monitoring. This study proposed a green remediation approach to treat and recycle the contaminated sediment by means of stabilization/solidification enhanced by the addition of selected solid wastes. With an increasing amount of contaminated sediment (20-70%), the 28-d compressive strength of sediment blocks decreased from greater than 10MPa to slightly above 1MPa. For augmenting the cement hydration, coal fly ash was more effective than lime and ground seashells, especially at low sediment content. The microscopic and spectroscopic analyses showed varying amounts of hydration products (primarily calcium hydroxide and calcium silicate hydrate) in the presence of coal fly ash, signifying the influence of pozzolanic reaction. To facilitate the waste utilization, cullet from beverage glass bottles and bottom ashes from coal combustion and waste incineration were found suitable to substitute coarse aggregate at 33% replacement ratio, beyond which the compressive strength decreased accordingly. The mercury intrusion porosimetry analysis indicated that the increase in the total pore area and average pore diameter were linearly correlated with the decrease of compressive strength due to waste replacement. All the sediment blocks complied with the acceptance criteria for reuse in terms of metal leachability. These results suggest that, with an appropriate mixture design, contaminated sediment and waste materials are useful resources for producing non-load-bearing masonry units or fill materials for construction uses. Copyright © 2014 Elsevier Ltd. All rights reserved.

  16. Disposal and improvement of contaminated by waste extraction of copper mining in chile

    Science.gov (United States)

    Naranjo Lamilla, Pedro; Blanco Fernández, David; Díaz González, Marcos; Robles Castillo, Marcelo; Decinti Weiss, Alejandra; Tapia Alvarez, Carolina; Pardo Fabregat, Francisco; Vidal, Manuel Miguel Jordan; Bech, Jaume; Roca, Nuria

    2016-04-01

    This project originated from the need of a mining company, which mines and processes copper ore. High purity copper is produced with an annual production of 1,113,928 tons of concentrate to a law of 32%. This mining company has generated several illegal landfills and has been forced by the government to make a management center Industrial Solid Waste (ISW). The forecast volume of waste generated is 20,000 tons / year. Chemical analysis established that the studied soil has a high copper content, caused by nature or from the spread of contaminants from mining activities. Moreover, in some sectors, soil contamination by mercury, hydrocarbons and oils and fats were detected, likely associated with the accumulation of waste. The waters are also impacted by mining industrial tasks, specifically copper ores, molybdenum, manganese, sulfates and have an acidic pH. The ISW management center dispels the pollution of soil and water and concentrating all activities in a technically suitable place. In this center the necessary guidelines for the treatment and disposal of soil contamination caused by uncontrolled landfills are given, also generating a leachate collection system and a network of fluid monitoring physicochemical water quality and soil environment. Keywords: Industrial solid waste, soil contamination, Mining waste

  17. Comprehensive implementation plan for the DOE defense buried TRU- contaminated waste program

    International Nuclear Information System (INIS)

    Everette, S.E.; Detamore, J.A.; Raudenbush, M.H.; Thieme, R.E.

    1988-02-01

    In 1970, the US Atomic Energy Commission established a ''transuranic'' (TRU) waste classification. Waste disposed of prior to the decision to retrievably store the waste and which may contain TRU contamination is referred to as ''buried transuranic-contaminated waste'' (BTW). The DOE reference plan for BTW, stated in the Defense Waste Management Plan, is to monitor it, to take such remedial actions as may be necessary, and to re-evaluate its safety as necessary or in about 10-year periods. Responsibility for management of radioactive waste and byproducts generated by DOE belongs to the Secretary of Energy. Regulatory control for these sites containing mixed waste is exercised by both DOE (radionuclides) and EPA (hazardous constituents). Each DOE Operations Office is responsible for developing and implementing plans for long-term management of its radioactive and hazardous waste sites. This comprehensive plan includes site-by-site long-range plans, site characteristics, site costs, and schedules at each site. 13 figs., 15 tabs

  18. Electrokinetic applications for environmental restoration, waste volume reduction, and contaminant containment systems

    International Nuclear Information System (INIS)

    Lomasney, H.L.; Lomasney, C.A.

    1996-01-01

    In the US and all over the world, following over 50 years of nuclear arms production operations, the magnitude of resultant environmental damage is only beginning to surface. The US Department of Energy estimates that by the year 2070, the total volume of high-level waste, transuranic waste, low-level waste, and low-level mixed waste, generated as a result of past and current nuclear activities, will exceed 20 million cubic meters. In Russia, it is reported that more than 30% of all groundwater is contaminated with agricultural and industrial chemical waste. Government agencies today are faced with the responsibility of developing technologies that are suitable for dealing with severe environmental contamination and accumulating waste inventories. In response to this demand, applications of electrokinetics have emerged in the field of environmental waste management as alternatives for environmental decontamination and ecological protection. Electrokinetics involves the movement of charged species under the influence of an applied electric field and is applicable in several areas of environmental waste management, including cleanup of soil and groundwater, barrier detection, and emergency or protective fencing. The worldwide interest in this technology has steadily escalated over the past decade. Today, state-of-the-art applications of electrokinetics have been demonstrated in the US, The Netherlands, Russia, The Ukraine, and India. This paper addresses the latest advances in the various applications of this technology as well as the most significant breakthroughs in the history of electrokinetics

  19. Risks from principal components and their daughter products in alpha-contaminated waste

    International Nuclear Information System (INIS)

    Rogers, V.

    1982-01-01

    This paper presents an overview on risk assessment, particularly as it applies to limits for alpha-contaminated waste. The general conclusions are: (1) no special characteristics of transuranic (TRU) waste justify its being a special category; (2) model calculations are largely subjective and can be influenced by the bias of the modeler; (3) ingrowth and concentration of TRU daughter products could be an important consideration in risk assessment. 13 figures

  20. Volume reduction technology of radioactive waste and clearance practice of contaminated material

    International Nuclear Information System (INIS)

    Gao Chao

    2016-01-01

    • One of principles of RW management is minimization and reduction: - Advance process and facilities should be reasonably applied to reduce the waste generation (''Law of the People's Republic of China on Prevention and Control of Radioactive Pollution'', 2003); - Operator of RW storage facilities should dispose or clear up solid waste timely (''Regulations on the safety of RW management'', 2011); • Reduction principle: - Control of generation; - Use of volume reduction technique; - Clearance of slightly contaminated material

  1. Theoretical considerations on design and analysis of monitoring systems for Pu-contaminated solid wastes

    International Nuclear Information System (INIS)

    Notea, A.

    1979-01-01

    Monitoring systems for plutonium contaminated wastes refers to both managerial regulations and instrumental hardware. Its design is inseparable from the design of the production line in the fuel handling facility, and depends on the general wastes management program. Characteristic functions of the monitors are discussed and the necessity of reference monitors is stressed. The reference monitor enables the formation of a quality scale. Guidelines for future R and D efforts are suggested

  2. Multiple containment for LSA [low specific activity] and SCO [surface contaminated objects] wastes

    International Nuclear Information System (INIS)

    Burgess, M.H.

    1993-09-01

    Radioactive wastes are generally transported in the form of Low Specific Activity (LSA) materials or Surface Contaminated Objects (SCO). This report proposes that a method of acknowledging the beneficial effects of multiple containment for such wastes should be written into the 1996 Edition of the IAEA Transport Regulations. Experience used to assess risks from on-site movements of radioactive material in the UK can be applied to develop safety arguments justifying the alleviation of off-site transport risks. (UK)

  3. Advances in measurement of alpha-contaminated wastes

    International Nuclear Information System (INIS)

    Close, D.A.; Crane, T.W.; Caldwell, J.T.; Kunz, W.E.; Shunk, E.R.; Pratt, J.C.; Franks, L.A.; Kominski, S.M.

    1982-01-01

    A comprehensive program is in progress at the Los Alamos National Laboratory for the development of sensitive, practical, nondestructive assay techniques for the quantification of low-level transuranics in bulk solid wastes. The program encompasses a broad range of techniques, including sophisticated active and passive gamma-ray spectroscopy, passive neutron detection systems, pulsed portable neutron generator interrogation systems, and electron accelerator-based techniques. The techniques can be used with either low-level or high-level beta-gamma wastes in either low-density or high-density matrices

  4. Stakeholder involvement in the evaluation of a multipurpose canister system

    International Nuclear Information System (INIS)

    Williams, J.R.; Kane, D.; Smith, T.B. Jr.

    1994-01-01

    The U.S. Department of Energy (DOE), Office of Civilian Radioactive Waste Management (OCRWM), began evaluating a multipurpose canister (MPC) concept in October of 1992. This followed recommendations by the Nuclear Waste Technical Review Board (NWTRB) and the U.S. Nuclear Regulatory Commission (NRC) that DOE develop a nuclear waste management system that achieves system integration, standardization, and reduced fuel-handling operations. Industry organizations such as Edison Electric Institute (EEI) and Electric Power Research Institute (EPRI) had conducted earlier studies that concluded advantages to the nuclear waste management system may be offered by such a concept. The MPC concept involves a metal canister which would contain multiple spent nuclear fuel assemblies. The canister would be sealed at the nuclear power plant and would not be reopened. The MPC would then be placed inside separate casks or overpacks for storage, transportation, and disposal. An important factor in DOE's evaluation of the MPC concept was the involvement of external parties. This paper describes that involvement process for the OCRWM's MPC implementation program. External parties who have an interest or stake in the program are referred to as stakeholders

  5. Contaminant Leach Testing of Hanford Tank 241-C-104 Residual Waste

    Energy Technology Data Exchange (ETDEWEB)

    Cantrell, Kirk J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Snyder, Michelle M.V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wang, Guohui [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buck, Edgar C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-07-01

    Leach testing of Tank C-104 residual waste was completed using batch and column experiments. Tank C-104 residual waste contains exceptionally high concentrations of uranium (i.e., as high as 115 mg/g or 11.5 wt.%). This study was conducted to provide data to develop contaminant release models for Tank C-104 residual waste and Tank C-104 residual waste that has been treated with lime to transform uranium in the waste to a highly insoluble calcium uranate (CaUO4) or similar phase. Three column leaching cases were investigated. In the first case, C-104 residual waste was leached with deionized water. In the second case, crushed grout was added to the column so that deionized water contacted the grout prior to contacting the waste. In the third case, lime was mixed in with the grout. Results of the column experiments demonstrate that addition of lime dramatically reduces the leachability of uranium from Tank C-104 residual waste. Initial indications suggest that CaUO4 or a similar highly insoluble calcium rich uranium phase forms as a result of the lime addition. Additional work is needed to definitively identify the uranium phases that occur in the as received waste and the waste after the lime treatment.

  6. Decontamination flowsheet development for a waste oil containing mixed radioactive contaminants

    International Nuclear Information System (INIS)

    Vijayan, S.; Buckley, L.P.

    1993-01-01

    The majority of waste oils contaminated with both radioactive and hazardous components are generated in nuclear power plant, research lab. and uranium-refinery operations. The waste oils are complex, requiring a detailed examination of the waste management strategies and technology options. It may appear that incineration offers a total solution, but this may not be true in all cases. An alternative approach is to decontaminate the waste oils to very low contaminant levels, so that the treated oils can be reused, burned as fuel in boilers, or disposed of by commercial incineration. This paper presents selected experimental data and evaluation results gathered during the development of a decontamination flowsheet for a specific waste oil stores at Chalk River Labs. (CRL). The waste oil contains varying amounts of lube oils, grease, paint, water, particulates, sludge, light chloro- and fluoro-solvents, polychlorinated biphenyls (PCB), complexing chemicals, uranium, chromium, iron, arsenic and manganese. To achieve safe management of this radioactive and hazardous waste, several treatment and disposal methods were screened. Key experiments were performed at the laboratory-scale to confirm and select the most appropriate waste-management scheme based on technical, environmental and economic criteria. The waste-oil-decontamination flowsheet uses a combination of unit operations, including prefiltration, acid scrubbing, and aqueous-leachage treatment by precipitation, microfiltration, filter pressing and carbon adsorption. The decontaminated oil containing open-quotes de minimisclose quotes levels of contaminants will undergo chemical destruction of PCBs and final disposal by incineration. The recovered uranium will be recycled to a uranium milling process

  7. Concerning enactment of regulations on burying of waste of nuclear fuel material or waste contaminated with nuclear fuel material

    International Nuclear Information System (INIS)

    1988-01-01

    The Atomic Safety Commission of Japan, after examining a report submitted by the Science and Technology Agency concerning the enactment of regulations on burying of waste of nuclear fuel material or waste contaminated with nuclear fuel material, has approved the plan given in the report. Thus, laws and regulations concerning procedures for application for waste burying business, technical standards for implementation of waste burying operation, and measures to be taken for security should be established to ensure the following. Matters to be described in the application for the approval of such business and materials to be attached to the application should be stipulated. Technical standards concerning inspection of waste burying operation should be stipulated. Measures to be taken for the security of waste burying facilities and security concerning the transportation and disposal of nuclear fuel material should be stipulated. Matters to be specified in the security rules should be stipulated. Matters to be recorded by waste burying business operators, measures to be taken to overcome dangers and matters to be reported to the Science and Technology Agency should be stipulated. (Nogami, K.)

  8. Household hazardous waste in municipal landfills: contaminants in leachate

    International Nuclear Information System (INIS)

    Slack, R.J.; Gronow, J.R.; Voulvoulis, N.

    2005-01-01

    Household hazardous waste (HHW) includes waste from a number of household products such as paint, garden pesticides, pharmaceuticals, photographic chemicals, certain detergents, personal care products, fluorescent tubes, waste oil, heavy metal-containing batteries, wood treated with dangerous substances, waste electronic and electrical equipment and discarded CFC-containing equipment. Data on the amounts of HHW discarded are very limited and are hampered by insufficient definitions of what constitutes HHW. Consequently, the risks associated with the disposal of HHW to landfill have not been fully elucidated. This work has focused on the assessment of data concerning the presence of hazardous chemicals in leachates as evidence of the disposal of HHW in municipal landfills. Evidence is sought from a number of sources on the occurrence in landfill leachates of hazardous components (heavy metals and xenobiotic organic compounds [XOC]) from household products and the possible disposal-to-emissions pathways occurring within landfills. This review demonstrates that a broad range of xenobiotic compounds occurring in leachate can be linked to HHW but further work is required to assess whether such compounds pose a risk to the environment and human health as a result of leakage/seepage or through treatment and discharge

  9. Summary of Preliminary Criticality Analysis for Peach Bottom Fuel in the DOE Standardized Spent Nuclear Fuel Canister

    International Nuclear Information System (INIS)

    Henrikson, D.J.

    1999-01-01

    The Department of Energy's (DOE's) National Spent Nuclear Fuel Program is developing a standardized set of canisters for DOE spent nuclear fuel (SNF). These canisters will be used for DOE SNF handling, interim storage, transportation, and disposal in the national repository. Several fuels are being examined in conjunction with the DOE SNF canisters. This report summarizes the preliminary criticality safety analysis that addresses general fissile loading limits for Peach Bottom graphite fuel in the DOE SNF canister. The canister is considered both alone and inside the 5-HLW/DOE Long Spent Fuel Co-disposal Waste Package, and in intact and degraded conditions. Results are appropriate for a single DOE SNF canister. Specific facilities, equipment, canister internal structures, and scenarios for handling, storage, and transportation have not yet been defined and are not evaluated in this analysis. The analysis assumes that the DOE SNF canister is designed so that it maintains reasonable geometric integrity. Parameters important to the results are the canister outer diameter, inner diameter, and wall thickness. These parameters are assumed to have nominal dimensions of 45.7-cm (18.0-in.), 43.815-cm (17.25-in), and 0.953-cm (0.375-in.), respectively. Based on the analysis results, the recommended fissile loading for the DOE SNF canister is 13 Peach Bottom fuel elements if no internal steel is present, and 15 Peach Bottom fuel elements if credit is taken for internal steel

  10. ISOCELL trademark proof-of-concept for retrieval of wastes and contaminated soil

    International Nuclear Information System (INIS)

    Chatwin, T.D.; Krieg, R.K.

    1992-01-01

    ISOCELL TM cryogenic technology is designed to immobilize buried hazardous, radioactive, and mixed waste and contaminated soil by creating a block of frozen waste and soil that can be safely retrieved, stored, transported, and treated with a minimum of dust or aerosol production. A ''proof-of-concept'' test of the ISOCELL process was conducted in clean soil by RKK, Ltd., for the Idaho National Engineering Laboratory (INEL). Results indicate ISOCELL technology successfully froze moist soil into a solid block capable of being lifted and retrieved. Test conditions were compared to characteristics of possible buried waste sites in the INEL

  11. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    Hamilton, L.

    2001-01-01

    This paper covers the design, development and testing of the magazines (cylinders containing cans of plutonium-ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a jointed robotic arm that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102mm) diameter canister throat

  12. The Biogeochemistry of Contaminant Groundwater Plumes Arising from Waste Disposal Facilities

    DEFF Research Database (Denmark)

    Bjerg, Poul Løgstrup; Albrechtsen, Hans-Jørgen; Kjeldsen, Peter

    2014-01-01

    Landfills with solid waste are abundant sources of groundwater pollution all over the world. Old uncontrolled municipal landfills are often large, heterogeneous sources with demolition waste, minor fractions of commercial or industrial waste, and organic waste from households. Strongly anaerobic...... leachate with a high content of dissolved organic carbon, salts, and ammonium, as well as specific organic compounds and metals is released from the waste for decades or centuries. Landfill leachate plume hosts a variety of biogeochemical processes, which is the key to understand the significant potential...... and the literature are the following: (1) Local hydrogeological conditions in the landfill area may affect the spreading of the contaminants; (2) investigations of landfill leachate plumes in geologic settings with clayey till deposits and fractured consolidated sediments are lacking; (3) the size of the landfill...

  13. Test design requirements: Canister-scale heater test

    International Nuclear Information System (INIS)

    Schauer, M.I.; Craig, P.A.; Stickney, R.G.

    1986-03-01

    This document establishes the Test Design Requirements for the design of a canister scale heater test to be performed in the Exploratory Shaft test facility. The purpose of the test is to obtain thermomechanical rock mass response data for use in validation of the numerical models. The canister scale heater test is a full scale simulation of a high-level nuclear waste container in a prototypic emplacement borehole. Electric heaters are used to simulate the heat loads expected in an actual waste container. This document presents an overview of the test including objectives and justification for the test. A description of the test as it is presently envisioned is included. Discussions on Quality Assurance and Safety are also included in the document. 12 refs., 1 fig

  14. Screening of contaminants in Waste Area Grouping 2 at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    Blaylock, B.G.; Frank, M.L.; Hoffman, F.O.; Hook, L.A.; Suter, G.W.; Watts, J.A.

    1992-07-01

    Waste Area Grouping 2 (WAG 2) of the Oak Ridge National Laboratory (ORNL) is located in the White Oak Creek Watershed and is composed of White Oak Creek Embayment, White Oak Lake and associated floodplain, and portions of White Oak Creek (WOC) and Melton Branch downstream of ORNL facilities. Contaminants leaving other ORNL WAGs in the WOC watershed pass through WAG 2 before entering the Clinch River. Health and ecological risk screening analyses were conducted on contaminants in WAG 2 to determine which contaminants were of concern and would require immediate consideration for remedial action and which contaminants could be assigned a low priority or further study. For screening purposes, WAG 2 was divided into four geographic reaches: Reach 1, a portion of WOC; Reach 2, Melton Branch; Reach 3, White Oak Lake and the floodplain area to the weirs on WOC and Melton Branch; and Reach 4, the White Oak Creek Embayment, for which an independent screening analysis has been completed. Screening analyses were conducted using data bases compiled from existing data on carcinogenic and noncarcinogenic contaminants, which included organics, inorganics, and radionuclides. Contaminants for which at least one ample had a concentration above the level of detection were placed in a detectable contaminants data base. Those contaminants for which all samples were below the level of detection were placed in a nondetectable contaminants data base

  15. EMERGING TECHNOLOGY SUMMARY: VITRIFICATION OF SOILS CONTAMINATED BY HAZARDOUS AND/OR RADIOACTIVE WASTES

    Science.gov (United States)

    A performance summary of an advanced multifuel-capable combustion and melting system (CMS) for the vitrification of hazardous wastes is presented. Vortex Corporation has evaluated its patented CMS for use in the remediation of soils contaminated with heavy metals and radionuclid...

  16. CHEMICAL MARKERS OF HUMAN WASTE CONTAMINATION IN SOURCE WATERS: A SIMPLIFIED ANALYTICAL APPROACH

    Science.gov (United States)

    Giving public water authorities a tool to monitor and measure levels of human waste contamination of waters simply and rapidly would enhance public protection. This methodology, using both urobilin and azithromycin (or any other human-use pharmaceutical) could be used to give pub...

  17. LAND TREATMENT AND THE TOXICITY RESPONSE OF SOIL CONTAMINATED WITH WOOD PRESERVING WASTE

    Science.gov (United States)

    Soils contaminated with wood preserving wastes, including pentachlo-rophenol (PCP) and creosote, are treated at field-scale in an engineered prepared-bed system consisting of two one-acre land treatment units (LTUs). The concentration of selected indicator compounds of treatment ...

  18. Processing results of 1,800 gallons of mercury and radioactively contaminated mixed waste rinse solution

    International Nuclear Information System (INIS)

    Thiesen, B.P.

    1993-01-01

    The mercury-contaminated rinse solution (INEL waste ID number-sign 123; File 8 waste) was successfully treated at the Idaho National Engineering Laboratory (INEL). This waste was generated during the decontamination of the Heat Transfer Reactor Experiment 3 (HTRE-3) reactor shield tank. Approximately 1,800 gal of waste was generated and was placed into 33 drums. Each drum contained precipitated sludge material ranging from 1--10 in. in depth, with the average depth of about 2.5 in. The pH of each drum varied from 3--11. The bulk liquid waste had a mercury level of 7.0 mg/l, which exceeded the Resource Conservation and Recovery Act (RCRA) limit of 0.2 mg/l. The average liquid bulk radioactivity was about 2.1 pCi/ml, while the average sludge contamination was about 13,800 pci/g. Treatment of the waste required separation of the liquid from the sludge, filtration, pH adjustment, and ion exchange. Because of difficulties in processing, three trials were required to reduce the mercury levels to below the RCRA limit. In the first trial, insufficient filtration of the waste allowed solid particulate produced during pH adjustment to enter into the ion exchange columns and ultimately the waste storage tank. In the second trial, the waste was filtered down to 0.1 μ to remove all solid mercury compounds. However, before filtration could take place, a solid mercury complex dissolved and mercury levels exceeded the RCRA limit after filtration. In the third trial, the waste was filtered through 0.3-A filters and then passed through the S-920 resin to remove the dissolved mercury. The resulting solut

  19. A welding system for spent fuel canister lid

    International Nuclear Information System (INIS)

    Suikki, M.; Wendelin, T.

    2008-06-01

    located in an adjacent auxiliary system room for facilitating maintenance and repair measures. The lid welding chamber has a low contamination risk but, during the presence of a canister in the welding chamber, the welding room must be kept restricted because of a high radiation level. The system is controlled from a fuel handling cell operation control room. Attached to this report is a separate research report, disclosing various heating calculations for the canister's copper overpack. Calculations have been conducted with regard to the effects on the copper overpack from preheating necessary for installing the lid, as well as with regard to the heating of a canister's shell during the actual encapsulation welding. The total cost estimate, without a value added tax, for manufacturing the apparatus amounted 1 551 000 euros, including 242 000 euros for designing costs and 151 000 euros for installation costs. (orig.)

  20. Study of a waste disposal site and it's groundwater contamination ...

    African Journals Online (AJOL)

    The choice of an old borrow pit at Avu village in the outskirts of Owerri Urban as the permanent dump for wastes from Owerri Urban is evaluated in terms of the hydrogeology of the site. The depth to the groundwater table or the vadose zone is 9 – 9.5m; the texture of the soils shows fine attenuative materials that can inhibit ...

  1. BIOREMEDIATION OF CONTAMINATED WASTE BY CADMIUM (Cd) IN WATERS USING INDIGEN BACTERIUM WITH EX-SITU WAY

    OpenAIRE

    Titik Wijayanti; Dinna Eka Graha Lestari

    2017-01-01

    The bioremediation technique for a contaminated liquid waste of heavy metals using indigenous bacteria is a convenient alternative to steps continues to be developed. The research aims to find out the effectiveness of an indigenous bacterial consortium in bioremediation of contaminated liquid waste by cadmium by ex-situ. Experiments were arranged in RAL made in ex-situ where a liquid waste industry was given five treatments, namely control and four indigenous bacterial consortia (A, D, E, and...

  2. Waste treatment process for removal of contaminants from aqueous, mixed-waste solutions using sequential chemical treatment and crossflow microfiltration, followed by dewatering

    Science.gov (United States)

    Vijayan, S.; Wong, C.F.; Buckley, L.P.

    1994-11-22

    In processes of this invention aqueous waste solutions containing a variety of mixed waste contaminants are treated to remove the contaminants by a sequential addition of chemicals and adsorption/ion exchange powdered materials to remove the contaminants including lead, cadmium, uranium, cesium-137, strontium-85/90, trichloroethylene and benzene, and impurities including iron and calcium. Staged conditioning of the waste solution produces a polydisperse system of size enlarged complexes of the contaminants in three distinct configurations: water-soluble metal complexes, insoluble metal precipitation complexes, and contaminant-bearing particles of ion exchange and adsorbent materials. The volume of the waste is reduced by separation of the polydisperse system by cross-flow microfiltration, followed by low-temperature evaporation and/or filter pressing. The water produced as filtrate is discharged if it meets a specified target water quality, or else the filtrate is recycled until the target is achieved. 1 fig.

  3. Chemical durability of copper canisters under crystalline bedrock repository conditions

    International Nuclear Information System (INIS)

    Sjoeblom, R.; Hermansson, H.P.; Amcoff, Oe.

    1995-01-01

    In the Swedish waste management programme, the copper canister is expected to provide containment of the radionuclides for a very long time, perhaps million of years. The purpose of the present paper is to analyze prerequisites for assessments of corrosion lifetimes for copper canisters. The analysis is based on compilations of literature from the following areas: chemical literature on copper and copper corrosion, mineralogical literature with emphasis on the stability of copper in near surface environments, and chemical and mineralogical literature with emphasis on the stabilities and thermodynamics of species and phases that may exist in a repository environment. Three main types of situations are identified: (1) under oxidizing and low chloride conditions, passivating oxide type of layers may form on the copper surface; (2) under oxidizing and high chloride conditions, the species formed may all be dissolved; and (3) under reducing conditions, non-passivating sulfide type layers may form on the copper surface. Considerable variability and uncertainty exists regarding the chemical environment for the canister, especially in certain scenarios. Thus, the mechanisms for corrosion can be expected to differ greatly for different situations. The lifetime of a thick-walled copper canister subjected to general corrosion appears to be long for most reasonable chemistries. Localized corrosion may appear for types (1) and (3) above but the mechanisms are widely different in character. The penetration caused by localized corrosion can be expected to be very sensitive to details in the chemistry. 20 refs, 3 figs, 1 tab

  4. Spatial assessment of soil contamination by heavy metals from informal electronic waste recycling in Agbogbloshie, Ghana.

    Science.gov (United States)

    Kyere, Vincent Nartey; Greve, Klaus; Atiemo, Sampson M

    2016-01-01

    This study examined the spatial distribution and the extent of soil contamination by heavy metals resulting from primitive, unconventional informal electronic waste recycling in the Agbogbloshie e-waste processing site (AEPS) in Ghana. A total of 132 samples were collected at 100 m intervals, with a handheld global position system used in taking the location data of the soil sample points. Observing all procedural and quality assurance measures, the samples were analyzed for barium (Ba), cadmium (Cd), cobalt (Co), chromium (Cr), copper (Cu), mercury (Hg), nickel (Ni), lead (Pb), and zinc (Zn), using X-ray fluorescence. Using environmental risk indices of contamination factor and degree of contamination (C deg ), we analyzed the individual contribution of each heavy metal contamination and the overall C deg . We further used geostatistical techniques of spatial autocorrelation and variability to examine spatial distribution and extent of heavy metal contamination. Results from soil analysis showed that heavy metal concentrations were significantly higher than the Canadian Environmental Protection Agency and Dutch environmental standards. In an increasing order, Pb>Cd>Hg>Cu>Zn>Cr>Co>Ba>Ni contributed significantly to the overall C deg . Contamination was highest in the main working areas of burning and dismantling sites, indicating the influence of recycling activities. Geostatistical analysis also revealed that heavy metal contamination spreads beyond the main working areas to residential, recreational, farming, and commercial areas. Our results show that the studied heavy metals are ubiquitous within AEPS and the significantly high concentration of these metals reflect the contamination factor and C deg , indicating soil contamination in AEPS with the nine heavy metals studied.

  5. Estimating risk at a Superfund site contaminated with radiological and chemical wastes

    International Nuclear Information System (INIS)

    Temeshy, A.; Liedle, J.M.; Sims, L.M.; Efird, C.R.

    1992-01-01

    This paper describes the method and results for estimating carcinogenic and noncarcinogenic effects at a Superfund site that is radiologically and chemically contaminated. Risk to receptors from disposal of waste in soil and resulting contamination of groundwater, air, surface water, and sediment is quantified. Specific risk assessment components which are addressed are the exposure assessment, toxicity assessment, and the resulting risk characterization. In the exposure assessment, potential exposure pathways are identified using waste disposal inventory information for soil and modeled information for other media. Models are used to calculate future radionuclide concentrations in groundwater, soil, surface water and air. Chemical exposure concentrations are quantified using site characterization data. Models are used to determine concentrations of chemicals in surface water and in air. Toxicity parameters used to quantify the dose-response relationship associated with the carcinogenic contaminants are slope factors and with noncarcinogenic contaminants are reference doses. In the risk characterization step, results from the exposure assessment and toxicity assessment are summarized and integrated into quantitative risk estimates for carcinogens and hazard induces for noncarcinogens. Calculated risks for carcinogenic contaminants are compared with EPA's target risk range. At WAG 6, the risk from radionuclides and chemicals for an on-WAG homesteader exceeds EPA's target risk range. Hazard indices are compared to unity for noncarcinogenic contaminants. At WAG 6, the total pathway hazard index for the on-WAG homesteader exceeds unity

  6. Performance of CASTORR HAW Cask Cold Trials for Loading, Transport and Storage of HAW canisters

    International Nuclear Information System (INIS)

    Wilmsmeier, Marco; Vossnacke, Andre

    2008-01-01

    On the basis of reprocessing contracts, concluded between the German Nuclear Utilities (GNUs) and the reprocessing companies in France (AREVA NC) and the UK (Nuclear Decommissioning Authority), GNS has the task to return the resulting residues to Germany. The high active waste (HAW) residuals from nuclear fuel reprocessing are vitrified and filled into steel cans, the HAW canisters. According to reprocessing contracts the equivalent number of HAW canisters to heavy metals delivered has to be returned to the country of origin and stored at an interim storage facility where applicable. The GNS' CASTOR R HAW casks are designed and licensed to fulfil the requirements for transport and long-term storage of HAW canisters. The new cask type CASTOR R HAW28M is capable of storing 28 HAW canisters with a maximum thermal power of 56 kW in total. Prior to the first active cask loading at a reprocessing facility it is required to demonstrate all important handling steps with the CASTOR R HAW28M cask according to a specific and approved sequence plan (MAP). These cold trials have to be carried out at the cask loading plant and at the reception area of an interim storage facility in Gorleben (TBL-G), witnessed by the licensing authorities and their independent experts. At transhipment stations GNS performs internal trials to demonstrate safe handling. A brand-new, empty CASTOR R HAW28M cask has been shipped from the GNS cask assembly facility in Muelheim to the TBL-G for cold trials. With this cask, GNS has to demonstrate the transhipment of casks at the Dannenberg transfer station from rail to road, transport to and reception at the TBL-G as well as incoming dose rate and contamination measurements and preparation for storage. After removal of all shock absorbers with a cask specific handling frame, tilting operation and assembly of the secondary lid with a pressure sensor, the helium leak tightness and 'Block-mass' tests have to be carried out as well. GNS long-term CASTOR R

  7. Treatment of Uranium-Contaminated Concrete for Reducing Secondary Radioactive Waste

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Soo; Han, G. S; Park, U. K; Kim, G. N.; Moon, J. K. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A volume reduction of the concrete waste by appropriate treatment technologies will decrease the amount of waste to be disposed of and result in a reduction of the disposal cost and an enhancement of the efficiency of the disposal site. Our group has developed a decontamination process for uranium-contaminated (U-contaminated) concrete, and some experiments were performed to reduce the second radioactive waste. A decontamination process was developed to remove uranium from concrete waste. The yellow or brown colored surface of the wall brick with high concentration of uranium was removed by a chisel until the radioactivity of remaining block reached less than 1 Bq/g. The concrete waste coated with epoxy was directly burned by an oil flame, and the burned surface was then removed using the same method as the treatment of the brick. The selective mechanical removal of the concrete block reduced the amount of secondary radioactive waste. The concrete blocks without an epoxy were crushed to below 30 mm and sifted to 1 mm. When the concrete pieces larger than 1 mm were sequentially washed with a clear recycle solution and 1.0 M of nitric acid, their radioactivity reached below the limit value of uranium for self-disposal. For the concrete pieces smaller than 1 mm, a rotary washing machine and electrokinetic equipment were also used.

  8. Treatment of Uranium-Contaminated Concrete for Reducing Secondary Radioactive Waste

    International Nuclear Information System (INIS)

    Kim, Seung Soo; Han, G. S; Park, U. K; Kim, G. N.; Moon, J. K.

    2014-01-01

    A volume reduction of the concrete waste by appropriate treatment technologies will decrease the amount of waste to be disposed of and result in a reduction of the disposal cost and an enhancement of the efficiency of the disposal site. Our group has developed a decontamination process for uranium-contaminated (U-contaminated) concrete, and some experiments were performed to reduce the second radioactive waste. A decontamination process was developed to remove uranium from concrete waste. The yellow or brown colored surface of the wall brick with high concentration of uranium was removed by a chisel until the radioactivity of remaining block reached less than 1 Bq/g. The concrete waste coated with epoxy was directly burned by an oil flame, and the burned surface was then removed using the same method as the treatment of the brick. The selective mechanical removal of the concrete block reduced the amount of secondary radioactive waste. The concrete blocks without an epoxy were crushed to below 30 mm and sifted to 1 mm. When the concrete pieces larger than 1 mm were sequentially washed with a clear recycle solution and 1.0 M of nitric acid, their radioactivity reached below the limit value of uranium for self-disposal. For the concrete pieces smaller than 1 mm, a rotary washing machine and electrokinetic equipment were also used

  9. Management of tritium contaminated wastes national strategies and practices at some European countries, USA and Canada

    International Nuclear Information System (INIS)

    Mannone, F.

    1992-01-01

    The European Tritium Handling Experiment Laboratory (ETHEL) is the Commission of European Communities facility designed for handling multigram quantities of tritium for safety inherent R and D purposes. Tritium contamined wastes in gaseous, liquid and solid forms will be generated in ETHEL during the experiments as well as during the maintenance operations. All such wastes must be adequately managed under the safest operating conditions to minimize the releases of tritium to the environment and the consequent radiological risks to workers and general population. This safety requirement can be met by carefully defining strategies and practices to be applied for the safe management of these wastes. To this end an adequate background information must be collected which is the intent of this report. Through an exhaustive literature survey current strategies and practices applied in Europe, USA and Canada for managing tritiated wastes from specific tritium handling laboratories and plant have been assessed. For some countries, where only tritium bearing wastes simultaneously contaminated with nuclear fission products are generated, the attention has been focused on the strategies and practices currently applied for managing fission wastes. Operational criteria for waste collection, sorting, classification, conditioning and packaging as well as acceptance criteria for their storage or disposal have been identified. Waste storage or disposal options already applied in various countries or still being investigated in terms of safety have also been considered. Even if the radwaste management strategy is submitted to a nearly continuing process of review, some general comments resulting from the assessment of the present waste management scenario are presented. 60 refs., 16 figs., 13 tabs

  10. Detection and Monitoring of E-Waste Contamination through Remote Sensing and Image Analysis

    Science.gov (United States)

    Garb, Yaakov; Friedlander, Lonia

    2015-04-01

    Electronic waste (e-waste) is one of today's fastest growing waste streams, and also one of the more problematic, as this end-of-life product contains precious metals mixed with and embedded in a variety of low value and potentially harmful plastic and other materials. This combination creates a powerful incentive for informal value chains that transport, extract from, and dispose of e-waste materials in far-ranging and unregulated ways, and especially in settings where regulation and livelihood alternatives are sparse, most notably in areas of India, China, and Africa. E-waste processing is known to release a variety of contaminants, such as heavy metals and persistent organic pollutants, including flame retardants, dioxins and furans. In several sites, where the livelihoods of entire communities are dependent on e-waste processing, the resulting contaminants have been demonstrated to enter the hydrological system and food chain and have serious health and ecological effects. In this paper we demonstrate for the first time the usefulness of multi-spectral remote sensing imagery to detect and monitor the release and possibly the dispersal of heavy metal contaminants released in e-waste processing. While similar techniques have been used for prospecting or for studying heavy metal contamination from mining and large industrial facilities, we suggest that these techniques are of particular value in detecting contamination from the more dispersed, shifting, and ad-hoc kinds of release typical of e-waste processing. Given the increased resolution and decreased price of multi-spectral imagery, such techniques may offer a remarkably cost-effective and rapidly responsive means of assessing and monitoring this kind of contamination. We will describe the geochemical and multi-spectral image-processing principles underlying our approach, and show how we have applied these to an area in which we have a detailed, multi-temporal, spatially referenced, and ground

  11. In-situ stabilization of mixed waste contaminated soil

    International Nuclear Information System (INIS)

    Siegrist, R.L.; Cline, S.R.; Gilliam, T.M.; Conner, J.R.

    1993-01-01

    A full-scale field demonstration was conducted to evaluate in for stabilizing an inactive RCRA land treatment site at a DOE facility in Ohio. Subsurface silt and clay deposits were contaminated principally with up to 500 mg/kg of trichloroethylene and other halocarbons, but also trace to low levels of Pb, Cr, 235 U, and 99 Tc. In situ solidification was studied in three, 3.1 m diameter by 4.6 m deep columns. During mixing, a cement-based grout was injected and any missions from the mixed region were captured in a shroud and treated by filtration and carbon adsorption. During in situ processing, operation and performance parameters were measured, and soil cores were obtained from a solidified column 15 months later. Despite previous site-specific treatability experience, there were difficulties in selecting a grout with the requisite treatment agents amenable to subsurface injection and at a volume adequate for distribution throughout the mixed region while minimizing volume expansion. observations during the demonstration revealed that in situ solidification was rapidly accomplished (e.g., >90 m 3 /d) with limited emissions of volatile organics (i.e., -6 cm/s vs. 10 -8 cm/s). Leaching tests performed on the treated samples revealed non-detectable to acceptably low concentrations of all target contaminants

  12. Simulation of heat transfer around a canister placed horizontally in a drift

    International Nuclear Information System (INIS)

    Moujaes, S.; Bhargava, A.

    1994-01-01

    The Yucca Mountain Site Characterization Project is investigating the feasibility of locating a high level radioactive nuclear waste repository at Yucca Mountain, Nevada. The bore hole and the in-drift waste emplacement schemes are under evaluation as potential repository drift geometries. This paper presents a two-dimensional finite element thermal analysis of the nuclear waste canister placed horizontally in a drift. Simulation has been carried out for 1000 years and the peak temperatures at the walls of the drift and at the center of the canister have been determined. The effect of the three modes of heat transfer, conduction, natural convection and radiation, is also discussed

  13. Radiological, physical, and chemical characterization of additional alpha contaminated and mixed low-level waste for treatment at the advanced mixed waste treatment project

    International Nuclear Information System (INIS)

    Hutchinson, D.P.

    1995-07-01

    This document provides physical, chemical, and radiological descriptive information for a portion of mixed waste that is potentially available for private sector treatment. The format and contents are designed to provide treatment vendors with preliminary information on the characteristics and properties for additional candidate portions of the Idaho National Engineering Laboratory (INEL) and offsite mixed wastes not covered in the two previous characterization reports for the INEL-stored low-level alpha-contaminated and transuranic wastes. This report defines the waste, provides background information, briefly reviews the requirements of the Federal Facility Compliance Act (P.L. 102-386), and relates the Site Treatment Plans developed under the Federal Facility Compliance Act to the waste streams described herein. Each waste is summarized in a Waste Profile Sheet with text, charts, and tables of waste descriptive information for a particular waste stream. A discussion of the availability and uncertainty of data for these waste streams precedes the characterization descriptions

  14. Radiological, physical, and chemical characterization of additional alpha contaminated and mixed low-level waste for treatment at the advanced mixed waste treatment project

    Energy Technology Data Exchange (ETDEWEB)

    Hutchinson, D.P.

    1995-07-01

    This document provides physical, chemical, and radiological descriptive information for a portion of mixed waste that is potentially available for private sector treatment. The format and contents are designed to provide treatment vendors with preliminary information on the characteristics and properties for additional candidate portions of the Idaho National Engineering Laboratory (INEL) and offsite mixed wastes not covered in the two previous characterization reports for the INEL-stored low-level alpha-contaminated and transuranic wastes. This report defines the waste, provides background information, briefly reviews the requirements of the Federal Facility Compliance Act (P.L. 102-386), and relates the Site Treatment Plans developed under the Federal Facility Compliance Act to the waste streams described herein. Each waste is summarized in a Waste Profile Sheet with text, charts, and tables of waste descriptive information for a particular waste stream. A discussion of the availability and uncertainty of data for these waste streams precedes the characterization descriptions.

  15. Dose potential of sludge contaminated and/or TRU contaminated waste in B-25s for tornado and straight wind events

    Energy Technology Data Exchange (ETDEWEB)

    Aponte, C.I.

    2000-02-17

    F and H Tank Farms generate supernate and sludge contaminated Low-Level Waste. The waste is collected, characterized, and packaged for disposal. Before the waste can be disposed of, however, it must be properly characterized. Since the radionuclide distribution in typical supernate is well known, its characterization is relatively straight forward and requires minimal effort. Non-routine waste, including potentially sludge contaminated, requires much more effort to effectively characterize. The radionuclide distribution must be determined. In some cases the waste can be contaminated by various sludge transfers with unique radionuclide distributions. In these cases, the characterization can require an extensive effort. Even after an extensive characterization effort, the container must still be prepared for shipping. Therefore a significant amount of time may elapse from the time the waste is generated until the time of disposal. During the time it is possible for a tornado or high wind scenario to occur. The purpose of this report is to determine the effect of a tornado on potential sludge contaminated waste, or Transuranic (TRU) waste in B-25s [large storage containers], to evaluate the potential impact on F and H Tank Farms, and to help establish a B-25 control program for tornado events.

  16. Dose potential of sludge contaminated and/or TRU contaminated waste in B-25s for tornado and straight wind events

    International Nuclear Information System (INIS)

    Aponte, C.I.

    2000-01-01

    F and H Tank Farms generate supernate and sludge contaminated Low-Level Waste. The waste is collected, characterized, and packaged for disposal. Before the waste can be disposed of, however, it must be properly characterized. Since the radionuclide distribution in typical supernate is well known, its characterization is relatively straight forward and requires minimal effort. Non-routine waste, including potentially sludge contaminated, requires much more effort to effectively characterize. The radionuclide distribution must be determined. In some cases the waste can be contaminated by various sludge transfers with unique radionuclide distributions. In these cases, the characterization can require an extensive effort. Even after an extensive characterization effort, the container must still be prepared for shipping. Therefore a significant amount of time may elapse from the time the waste is generated until the time of disposal. During the time it is possible for a tornado or high wind scenario to occur. The purpose of this report is to determine the effect of a tornado on potential sludge contaminated waste, or Transuranic (TRU) waste in B-25s [large storage containers], to evaluate the potential impact on F and H Tank Farms, and to help establish a B-25 control program for tornado events

  17. Carbon speciation in ash, residual waste and contaminated soil by thermal and chemical analyses.

    Science.gov (United States)

    Kumpiene, Jurate; Robinson, Ryan; Brännvall, Evelina; Nordmark, Désirée; Bjurström, Henrik; Andreas, Lale; Lagerkvist, Anders; Ecke, Holger

    2011-01-01

    Carbon in waste can occur as inorganic (IC), organic (OC) and elemental carbon (EC) each having distinct chemical properties and possible environmental effects. In this study, carbon speciation was performed using thermogravimetric analysis (TGA), chemical degradation tests and the standard total organic carbon (TOC) measurement procedures in three types of waste materials (bottom ash, residual waste and contaminated soil). Over 50% of the total carbon (TC) in all studied materials (72% in ash and residual waste, and 59% in soil) was biologically non-reactive or EC as determined by thermogravimetric analyses. The speciation of TOC by chemical degradation also showed a presence of a non-degradable C fraction in all materials (60% of TOC in ash, 30% in residual waste and 13% in soil), though in smaller amounts than those determined by TGA. In principle, chemical degradation method can give an indication of the presence of potentially inert C in various waste materials, while TGA is a more precise technique for C speciation, given that waste-specific method adjustments are made. The standard TOC measurement yields exaggerated estimates of organic carbon and may therefore overestimate the potential environmental impacts (e.g. landfill gas generation) of waste materials in a landfill environment. Copyright © 2010 Elsevier Ltd. All rights reserved.

  18. Characterization of projected DWPF glasses heat treated to simulate canister centerline cooling

    International Nuclear Information System (INIS)

    Marra, S.L.; Jantzen, C.M.

    1992-05-01

    Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced and poured into stainless steel canisters in the Defense Waste Processing Facility (DWPF). Eventually these canistered waste forms will be sent to a geologic repository for final disposal. In order to assure acceptability by the repository, the Department of Energy has defined requirements which DWPF canistered waste forms must meet. These requirements are the Waste Acceptance Preliminary Specifications (WAPS). The WAPS require DWPF to identify the crystalline phases expected to be present in the final glass product. Knowledge of the thermal history of the borosilicate glass during filling and cooldown of the canister is necessary to determine the amount and type of crystalline phases present in the final glass product. Glass samples of seven projected DWPF compositions were cooled following the same temperature profile as that of glass at the centerline of the full-scale DWPF canister. The glasses were characterized by x-ray diffraction and scanning electron microscopy to identify the crystalline phases present The volume percents of each crystalline phase present were determined by quantitative x-ray diffraction. The Product Consistency Test (PCI) was used to determine the durability of the heat-treated glasses

  19. Long-range alpha detection applied to soil contamination and waste monitoring

    International Nuclear Information System (INIS)

    MacArthur, D.W.; Allander, K.S.; Bounds, J.A.; Close, D.A.; McAtee, J.L.

    1992-01-01

    Alpha contamination monitoring has been traditionally limited by the short range of alpha particles in air and through detector windows. The long-range alpha detector (LRAD) described in this paper circumvents that limitation by detecting alpha-produced ions, rather than alpha particles directly. Since the LRAD is sensitive to all ions, it can monitor all contamination present on a large surface at one time. Because air is the ''detector gas,'' the LRAD can detect contamination on any surface to which air can penetrate. We present data showing the sensitivity of LRAD detectors, as well as documenting their ability to detect alpha sources in previously unmonitorable locations, and verifying the ion lifetime. Specific designs and results for soil contamination and waste monitors are also included

  20. Multi-isotopic gamma-ray assay system for alpha-contaminated waste

    International Nuclear Information System (INIS)

    Close, D.A.; Pratt, J.C.; Caldwell, J.T.; Kunz, W.E.; Schultz, F.J.; Haff, K.W.

    1983-01-01

    The capability of an existing segmented gamma-ray system is being expanded for the analysis of alpha-contaminated waste drums. A cursory assay of 114 transuranic waste drums of 208-l capacity has been made. Analysis of these data indicates a detection limit better than 100 nCi/g of waste for 237 Np/ 233 Pa, 239 Pu, 241 Am, 243 Am/ 239 Np, 60 Co, 125 Sb, 134 137 Cs, and 154 Eu. A pending Code of Federal Regulation (10CFR61) stipulates that the nuclear industry quantify not only its transuranic waste, but also certain beta- and gamma-ray-emitting fission products. An assay system based on gamma-ray spectroscopy is the only system that can meet this requirement for the fission products

  1. Solid waste leach characteristics and contaminant-sediment interactions Volume 2: Contaminant transport under unsaturated moisture contents

    International Nuclear Information System (INIS)

    Lindenmeier, C.W.; Serne, R.J.; Conca, J.L.

    1995-09-01

    The objectives of this report and subsequent volumes include describing progress on (1) development and optimization of experimental methods to quantify the release of contaminants from solid wastes and their subsequent interactions with unsaturated sediments and (2) the creation of empirical data that become input parameters to performance assessment (PA) analyses for future Hanford Site disposal units and baseline risk assessments for inactive and existing solid waste disposal units. For this report, efforts focused on developing methodologies to evaluate contaminant transport in Trench 8 (W-5 Burial Ground) sediments under unsaturated (vadose zone) conditions. To accomplish this task, a series of flow-through column tests were run using standard saturated column systems, Wierenga unsaturated column systems (both commercial and modified), and the Unsaturated Flow Apparatus (UFA). The reactants investigated were 85 Sr, 236 U, and 238 U as reactive tracers, and tritium as a non-reactive tracer. Results indicate that for moderately unsaturated conditions (volumetric water contents >50 % of saturation), the Wierenga system performed reasonably well such that long water residence times (50-147 h) were achieved, and reasonably good steady-state flow conditions were maintained. The major drawbacks in using this system for reactive tracer work included (1) the inability to achieve reproducible and constant moisture content below 50% of saturation, (2) the four to six month time required to complete a single test, and (3) the propensity for mechanical failure resulting from laboratory power outages during the prolonged testing period

  2. Bioremediation of oil-contaminated soil using Candida catenulata and food waste

    International Nuclear Information System (INIS)

    Joo, Hung-Soo; Ndegwa, Pius M.; Shoda, Makoto; Phae, Chae-Gun

    2008-01-01

    Even though petroleum-degrading microorganisms are widely distributed in soil and water, they may not be present in sufficient numbers to achieve contaminant remediation. In such cases, it may be useful to inoculate the polluted area with highly effective petroleum-degrading microbial strains to augment the exiting ones. In order to identify a microbial strain for bioaugmentation of oil-contaminated soil, we isolated a microbial strain with high emulsification and petroleum hydrocarbon degradation efficiency of diesel fuel in culture. The efficacy of the isolated microbial strain, identified as Candida catenulata CM1, was further evaluated during composting of a mixture containing 23% food waste and 77% diesel-contaminated soil including 2% (w/w) diesel. After 13 days of composting, 84% of the initial petroleum hydrocarbon was degraded in composting mixes containing a powdered form of CM1 (CM1-solid), compared with 48% of removal ratio in control reactor without inoculum. This finding suggests that CM1 is a viable microbial strain for bioremediation of oil-contaminated soil with food waste through composting processes. - Enhancement on degradation ability of petroleum hydrocarbon by the microbial strain in the composting process with food waste

  3. Organic Contaminant Content and Physico-Chemical Characteristics of Waste Materials Recycled in Agriculture

    Directory of Open Access Journals (Sweden)

    Hannah Rigby

    2015-12-01

    Full Text Available A range of wastes representative of materials currently applied, or with future potential to be applied, to agricultural land in the UK as fertilisers and soil improvers or used as animal bedding in livestock production, were investigated. In addition to full physico-chemical characterization, the materials were analysed for a suite of priority organic contaminants. In general, contaminants were present at relatively low concentrations. For example, for biosolids and compost-like-output (CLO, concentrations of polychlorinated dibenzo-p-dioxins/dibenzofurans (PCDD/Fs and polychlorinated biphenyls (PCBs were approximately 1−10 and 5–50 times lower, respectively, than various proposed or implemented European limit values for these contaminants in biosolids or composts applied to agricultural land. However, the technical basis for these limits may require re-evaluation in some cases. Polybrominated, and mixed halogenated, dibenzo-p-dioxins/dibenzofurans are not currently considered in risk assessments of dioxins and dioxin-like chemicals, but were detected at relatively high concentrations compared with PCDD/Fs in the biosolids and CLOs and their potential contribution to the overall toxic equivalency is assessed. Other ‘emerging’ contaminants, such as organophosphate flame retardants, were detected in several of the waste materials, and their potential significance is discussed. The study is part of a wider research programme that will provide evidence that is expected to improve confidence in the use of waste-derived materials in agriculture and to establish guidelines to protect the food chain where necessary.

  4. Potential for effects of land contamination on human health. 2. The case of waste disposal sites.

    Science.gov (United States)

    Kah, Melanie; Levy, Len; Brown, Colin

    2012-01-01

    This review of the epidemiological literature shows that evidence for negative impacts of land contaminated by waste disposal on human health is limited. However, the potential for health impacts cannot be dismissed. The link between residence close to hazardous waste disposal sites and heightened levels of stress and anxiety is relatively well established. However, studies on self-reported outcomes generally suffer from interpretational problems, as subjective symptoms may be due to increased perception and recall. Several recent multiple-site studies support a plausible linkage between residence near waste disposal sites and reproductive effects (including congenital anomalies and low birth weight). There is some conflict in the literature investigating links between land contamination and cancers; the evidence for and against a link is equally balanced and is insufficient to make causal inferences. These are difficult to establish because of lack of data on individual exposures, and other socioeconomic and lifestyle factors that may confound a relationship with area of residence. There is no consistently occurring risk for any specific tumor across multiple studies on sites expected to contain similar contaminants. Further insights on health effects of land contamination are likely to be gained from studies that consider exposure pathways and biomarkers of exposure and effect, similar to those deployed with some success in investigating impacts of cadmium on human health.

  5. Establishing indices for groundwater contamination risk assessment in the vicinity of hazardous waste landfills in China

    International Nuclear Information System (INIS)

    Li Ying; Li Jinhui; Chen Shusheng; Diao Weihua

    2012-01-01

    Groundwater contamination by leachate is the most damaging environmental impact over the entire life of a hazardous waste landfill (HWL). With the number of HWL facilities in China rapidly increasing, and considering the poor status of environmental risk management, it is imperative that effective environmental risk management methods be implemented. A risk assessment indices system for HWL groundwater contamination is here proposed, which can simplify the risk assessment procedure and make it more user-friendly. The assessment framework and indices were drawn from five aspects: source term, underground media, leachate properties, risk receptors and landfill management quality, and a risk assessment indices system consisting of 38 cardinal indicators was established. Comparison with multimedia models revealed that the proposed indices system was integrated and quantitative, that input data for it could be easily collected, and that it could be widely used for environmental risk assessment (ERA) in China. - Highlights: ► No comprehensive environmental risk assessment method for hazardous waste management is proposed in China. ► An assessment indices system is established for groundwater contamination in the vicinity of hazardous waste landfill. ► All indicators are quantitative and applicable in China. - Capsule: This research identified critical indices and established a system for environmental risk assessment for groundwater contamination in the vicinity of HWLs in China.

  6. Toward identifying the next generation of superfund and hazardous waste site contaminants.

    Science.gov (United States)

    Ela, Wendell P; Sedlak, David L; Barlaz, Morton A; Henry, Heather F; Muir, Derek C G; Swackhamer, Deborah L; Weber, Eric J; Arnold, Robert G; Ferguson, P Lee; Field, Jennifer A; Furlong, Edward T; Giesy, John P; Halden, Rolf U; Henry, Tala; Hites, Ronald A; Hornbuckle, Keri C; Howard, Philip H; Luthy, Richard G; Meyer, Anita K; Sáez, A Eduardo; Vom Saal, Frederick S; Vulpe, Chris D; Wiesner, Mark R

    2011-01-01

    This commentary evolved from a workshop sponsored by the National Institute of Environmental Health Sciences titled "Superfund Contaminants: The Next Generation" held in Tucson, Arizona, in August 2009. All the authors were workshop participants. Our aim was to initiate a dynamic, adaptable process for identifying contaminants of emerging concern (CECs) that are likely to be found in future hazardous waste sites, and to identify the gaps in primary research that cause uncertainty in determining future hazardous waste site contaminants. Superfund-relevant CECs can be characterized by specific attributes: They are persistent, bioaccumulative, toxic, occur in large quantities, and have localized accumulation with a likelihood of exposure. Although still under development and incompletely applied, methods to quantify these attributes can assist in winnowing down the list of candidates from the universe of potential CECs. Unfortunately, significant research gaps exist in detection and quantification, environmental fate and transport, health and risk assessment, and site exploration and remediation for CECs. Addressing these gaps is prerequisite to a preventive approach to generating and managing hazardous waste sites. A need exists for a carefully considered and orchestrated expansion of programmatic and research efforts to identify, evaluate, and manage CECs of hazardous waste site relevance, including developing an evolving list of priority CECs, intensifying the identification and monitoring of likely sites of present or future accumulation of CECs, and implementing efforts that focus on a holistic approach to prevention.

  7. Impacts of biochar and oyster shells waste on the immobilization of arsenic in highly contaminated soils.

    Science.gov (United States)

    Chen, Yongshan; Xu, Jinghua; Lv, Zhengyong; Xie, Ruijia; Huang, Liumei; Jiang, Jinping

    2018-07-01

    Soil contamination is a serious problem with deleterious impacts on global sustainability. Readily available, economic, and highly effective technologies are therefore urgently needed for the rehabilitation of contaminated sites. In this study, two readily available materials prepared from bio-wastes, namely biochar and oyster shell waste, were evaluated as soil amendments to immobilize arsenic in a highly As-contaminated soil (up to 15,000 mgAs/kg). Both biochar and oyster shell waste can effectively reduce arsenic leachability in acid soils. After application of the amendments (2-4% addition, w/w), the exchangeable arsenic fraction decreased from 105.8 to 54.0 mg/kg. The application of 2%biochar +2% oyster shell waste most effectively reduced As levels in the column leaching test by reducing the arsenic concentration in the porewater by 62.3% compared with the treatment without amendments. Biochar and oyster shell waste also reduced soluble As(III) from 374.9 ± 18.8 μg/L to 185.9 ± 16.8 μg/L and As(V) from 119.8 ± 13.0 μg/L to 56.4 ± 2.6 μg/L at a pH value of 4-5. The treatment using 4% (w/w) amendments did not result in sufficient As immobilization in highly contaminated soils; high soluble arsenic concentrations (upto193.0 μg/L)were found in the soil leachate, particularly in the form of As(III), indicating a significant potential to pollute shallow groundwater aquifers. This study provides valuable insights into the use of cost-effective and readily available materials for soil remediation and investigates the mechanisms underlying arsenic immobilization in acidic soils. Copyright © 2018 Elsevier Ltd. All rights reserved.

  8. Spent fuel canister docking station

    International Nuclear Information System (INIS)

    Suikki, M.

    2006-01-01

    The working report for the spent fuel canister docking station presents a design for the operation and structure of the docking equipment located in the fuel handling cell for the spent fuel in the encapsulation plant. The report contains a description of the basic requirements for the docking station equipment and their implementation, the operation of the equipment, maintenance and a cost estimate. In the designing of the equipment all the problems related with the operation have been solved at the level of principle, nevertheless, detailed designing and the selection of final components have not yet been carried out. In case of defects and failures, solutions have been considered for postulated problems, and furthermore, the entire equipment was gone through by the means of systematic risk analysis (PFMEA). During the docking station designing we came across with needs to influence the structure of the actual disposal canister for spent nuclear fuel, too. Proposed changes for the structure of the steel lid fastening screw were included in the report. The report also contains a description of installation with the fuel handling cell structures. The purpose of the docking station for the fuel handling cell is to position and to seal the disposal canister for spent nuclear fuel into a penetration located on the cell floor and to provide suitable means for executing the loading of the disposal canister and the changing of atmosphere. The designed docking station consists of a docking ring, a covering hatch, a protective cone and an atmosphere-changing cap as well as the vacuum technology pertaining to the changing of atmosphere and the inert gas system. As far as the solutions are concerned, we have arrived at rather simple structures and most of the actuators of the system are situated outside of the actual fuel handling cell. When necessary, the equipment can also be used for the dismantling of a faulty disposal canister, cut from its upper end by machining. The

  9. The development of a Martian atmospheric Sample collection canister

    Science.gov (United States)

    Kulczycki, E.; Galey, C.; Kennedy, B.; Budney, C.; Bame, D.; Van Schilfgaarde, R.; Aisen, N.; Townsend, J.; Younse, P.; Piacentine, J.

    The collection of an atmospheric sample from Mars would provide significant insight to the understanding of the elemental composition and sub-surface out-gassing rates of noble gases. A team of engineers at the Jet Propulsion Laboratory (JPL), California Institute of Technology have developed an atmospheric sample collection canister for Martian application. The engineering strategy has two basic elements: first, to collect two separately sealed 50 cubic centimeter unpressurized atmospheric samples with minimal sensing and actuation in a self contained pressure vessel; and second, to package this atmospheric sample canister in such a way that it can be easily integrated into the orbiting sample capsule for collection and return to Earth. Sample collection and integrity are demonstrated by emulating the atmospheric collection portion of the Mars Sample Return mission on a compressed timeline. The test results achieved by varying the pressure inside of a thermal vacuum chamber while opening and closing the valve on the sample canister at Mars ambient pressure. A commercial off-the-shelf medical grade micro-valve is utilized in the first iteration of this design to enable rapid testing of the system. The valve has been independently leak tested at JPL to quantify and separate the leak rates associated with the canister. The results are factored in to an overall system design that quantifies mass, power, and sensing requirements for a Martian atmospheric Sample Collection (MASC) canister as outlined in the Mars Sample Return mission profile. Qualitative results include the selection of materials to minimize sample contamination, preliminary science requirements, priorities in sample composition, flight valve selection criteria, a storyboard from sample collection to loading in the orbiting sample capsule, and contributions to maintaining “ Earth” clean exterior surfaces on the orbiting sample capsule.

  10. Testing contamination risk assessment methods for toxic elements from mine waste sites

    Science.gov (United States)

    Abdaal, A.; Jordan, G.; Szilassi, P.; Kiss, J.; Detzky, G.

    2012-04-01

    Major incidents involving mine waste facilities and poor environmental management practices have left a legacy of thousands of contaminated sites like in the historic mining areas in the Carpathian Basin. Associated environmental risks have triggered the development of new EU environmental legislation to prevent and minimize the effects of such incidents. The Mine Waste Directive requires the risk-based inventory of all mine waste sites in Europe by May 2012. In order to address the mining problems a standard risk-based Pre-selection protocol has been developed by the EU Commission. This paper discusses the heavy metal contamination in acid mine drainage (AMD) for risk assessment (RA) along the Source-Pathway-Receptor chain using decision support methods which are intended to aid national and regional organizations in the inventory and assessment of potentially contaminated mine waste sites. Several recognized methods such as the European Environmental Agency (EEA) standard PRAMS model for soil contamination, US EPA-based AIMSS and Irish HMS-IRC models for RA of abandoned sites are reviewed, compared and tested for the mining waste environment. In total 145 ore mine waste sites have been selected for scientific testing using the EU Pre-selection protocol as a case study from Hungary. The proportion of uncertain to certain responses for a site and for the total number of sites may give an insight of specific and overall uncertainty in the data we use. The Pre-selection questions are efficiently linked to a GIS system as database inquiries using digital spatial data to directly generate answers. Key parameters such as distance to the nearest surface and ground water bodies, to settlements and protected areas are calculated and statistically evaluated using STATGRAPHICS® in order to calibrate the RA models. According to our scientific research results, of the 145 sites 11 sites are the most risky having foundation slope >20o, 57 sites are within distance 66 (class VI

  11. [Simulation on contamination forecast and control of groundwater in a certain hazardous waste landfill].

    Science.gov (United States)

    Ma, Zhi-Fei; An, Da; Jiang, Yong-Hai; Xi, Bei-Dou; Li, Ding-Long; Zhang, Jin-Bao; Yang, Yu

    2012-01-01

    On the basis of site investigation and data collection of a certain hazardous waste landfill, the groundwater flow and solute transport coupled models were established by applying Visual Modflow software, which was used to conduct a numerical simulation that forecast the transport process of Cr6+ in groundwater and the effects of three control measures (ground-harden, leakage-proof barriers and drainage ditches) of contaminants transport after leachate leakage happened in impermeable layer of the landfill. The results show that the contamination plume of Cr6+ transports with groundwater flow direction, the contamination rang would reach the pool's boundary in 10 years, and the distance of contamination transport is 1 450 m. But the diffusion range of contamination plume would not be obviously expanded between 10 and 20 years. While the ground is hardened, the contamination plume would not reach the pool's boundary in 20 years. When the leakage-proof barrier is set in the bottom of water table aquifer, the concentration of Cr6+ is higher than that the leakage-proof barrier is unset, but the result is just opposite when setting the leakage-proof barrier in the bottom of underlying aquifer. The range of contamination plume is effectively controlled by setting drainage ditches that water discharge is 2 642 m3 x d(-1), which makes the monitoring wells would not be contaminated in 20 years. Moreover, combining the ground-harden with drainage ditches can get the best effect in controlling contaminants diffusion, and meanwhile, the drainage ditches' daily discharge is reduced to 1 878 m3 x d(-1). Therefore, it is suggested that the control measure combining the ground-harden with drainage ditches should apply to prevent contamination diffusion in groundwater when leachate leakage have happened in impermeable layer of the landfill.

  12. RECOVERY OF MERCURY FROM CONTAMINATED PRIMARY AND SECONDARY WASTES

    International Nuclear Information System (INIS)

    A. Faucette; J. Bognar; T. Broderick; T. Battaglia

    2000-01-01

    Effective removal of mercury contamination from water is a complex and difficult problem. In particular, mercury treatment of natural waters is difficult because of the low regulatory standards. For example, the Environmental Protection Agency has established a national ambient water quality standard of 12 parts-per-trillion (ppt), whereas the standard is 1.8 ppt in the Great Lakes Region. In addition, mercury is typically present in several different forms, but sorption processes are rarely effective with more than one or two of these forms. To meet the low regulatory discharge limits, a sorption process must be able to address all forms of mercury present in the water. One approach is to apply different sorbents in series depending on the mercury speciation and the regulatory discharge limits. Four new sorbents have been developed to address the variety of mercury species present in industrial discharges and natural waters. Three of these sorbents have been field tested on contaminated creek water at the Y-12 Plant. Two of these sorbents have demonstrated very high removal efficiencies for soluble mercury species, with mercury concentrations at the outlet of a pilot-scale system less than 12 ppt for as long as six months. The other sorbent tested at the Y-12 Plant is targeted at colloidal mercury that is not removed by standard sorption or filtration processes. At the Y-12 Plant, colloidal mercury appears to be associated with iron, so a sorbent that removes mercury-iron complexes in the presence of a magnetic field was evaluated. Field results indicate good removal of this mercury fraction from the Y-12 waters. In addition, this sorbent is easily regenerated by simply removing the magnetic field and flushing the columns with water. The fourth sorbent is still undergoing laboratory development, but results to date indicate exceptionally high mercury sorption capacity. The sorbent is capable of removing all forms of mercury typically present in natural and

  13. 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.)

  14. Environmental safety of the disposal system for radioactive substance-contaminated wastes

    International Nuclear Information System (INIS)

    Oosako, Masahiro

    2012-01-01

    In accordance with the full-scale enforcement of 'The Act on Special Measures concerning the Handling of Radioactive Pollution' in 2012, the collective efforts of entire Japan for dealing with radioactive pollutants began. The most important item for dealing with radioactive pollution is to control radioactive substances that polluted the global environment and establish a contaminated waste treatment system for risk reduction. On the incineration system and landfill disposal system of radioactive waste, this paper arranges the scientific information up to now, and discusses the safety of the treatment / disposal systems of contaminated waste. As for 'The Act on Special Measures concerning the Handling of Radioactive Pollution,' this paper discusses the points of the Act and basic policy, roadmap for the installation of interim storage facilities, and enforcement regulations (Ordinance of the Ministry of the Environment). About the safety of waste treatment system, it discusses the safety level of technical standards at waste treatment facilities, safety of incineration facilities, and safety of landfill disposal sites. (O.A.)

  15. SULFUR POLYMER STABILIZATION/SOLIDIFICATION (SPSS) TREATABILITY OF SIMULATED MIXED-WASTE MERCURY CONTAMINATED SLUDGE.

    Energy Technology Data Exchange (ETDEWEB)

    ADAMA, J.W.; BOWERMAN, B.S.; KALB, P.D.

    2002-10-01

    The Environmental Protection Agency (EPA) is currently seeking to validate technologies that can directly treat radioactively contaminated high mercury (Hg) subcategory wastes without removing the mercury from the waste. The Sulfur Polymer Stabilization/Solidification (SPSS) process developed at Brookhaven National Laboratory is one of several candidate technologies capable of successfully treating various Hg waste streams. To supplement previously supplied data on treatment of soils, EPA needs additional data concerning stabilization of high Hg subcategory waste sludges. To this end, a 5000 ppm sludge surrogate, containing approximately 50 wt% water, was successfully treated by pilot-scale SPSS processing. In two process runs, 85 and 95 wt% of water was recovered from the sludge during processing. At waste loadings of 30 wt% dry sludge, the treated waste form had no detectable mercury (<10 ppb) in TCLP leachates. Data gathered from the demonstration of treatment of this sludge will provide EPA with information to support revisions to current treatment requirements for high Hg subcategory wastes.

  16. SULFUR POLYMER STABILIZATION/SOLIDIFICATION (SPSS) TREATABILITY OF SIMULATED MIXED-WASTE MERCURY CONTAMINATED SLUDGE

    Energy Technology Data Exchange (ETDEWEB)

    Adams, J. W.; Bowerman, B. S.; Kalb, P. D.

    2002-02-25

    The Environmental Protection Agency (EPA) is currently evaluating alternative treatment standards for radioactively contaminated high mercury (Hg) subcategory wastes, which do not require the removal of mercury from the waste. The Sulfur Polymer Stabilization/Solidification (SPSS) process developed at Brookhaven National Laboratory is one of several candidate technologies capable of successfully treating various Hg waste streams. To supplement previously supplied data on treatment of soils, EPA needed additional data concerning stabilization of high Hg subcategory waste sludges. To this end, a 5000 ppm sludge surrogate, containing approximately 50 wt% water, was successfully treated by pilot-scale SPSS processing. In two process runs, 85 and 95 wt% of water was recovered from the sludge during processing. At waste loadings of 46 wt% (30 wt% dry) sludge, the treated waste form had no detectable mercury (<10 ppb) in TCLP leachates. Data gathered from the demonstration of treatment of this sludge will provide the EPA with information to support revisions to current treatment requirements for high Hg subcategory wastes.

  17. SULFUR POLYMER STABILIZATION/SOLIDIFICATION (SPSS) TREATABILITY OF SIMULATED MIXED-WASTE MERCURY CONTAMINATED SLUDGE

    International Nuclear Information System (INIS)

    ADAMA, J.W.; BOWERMAN, B.S.; KALB, P.D.

    2002-01-01

    The Environmental Protection Agency (EPA) is currently seeking to validate technologies that can directly treat radioactively contaminated high mercury (Hg) subcategory wastes without removing the mercury from the waste. The Sulfur Polymer Stabilization/Solidification (SPSS) process developed at Brookhaven National Laboratory is one of several candidate technologies capable of successfully treating various Hg waste streams. To supplement previously supplied data on treatment of soils, EPA needs additional data concerning stabilization of high Hg subcategory waste sludges. To this end, a 5000 ppm sludge surrogate, containing approximately 50 wt% water, was successfully treated by pilot-scale SPSS processing. In two process runs, 85 and 95 wt% of water was recovered from the sludge during processing. At waste loadings of 30 wt% dry sludge, the treated waste form had no detectable mercury (<10 ppb) in TCLP leachates. Data gathered from the demonstration of treatment of this sludge will provide EPA with information to support revisions to current treatment requirements for high Hg subcategory wastes

  18. SULFUR POLYMER STABILIZATION/SOLIDIFICATION (SPSS) TREATABILITY OF SIMULATED MIXED-WASTE MERCURY CONTAMINATED SLUDGE

    International Nuclear Information System (INIS)

    Adams, J. W.; Bowerman, B. S.; Kalb, P. D.

    2002-01-01

    The Environmental Protection Agency (EPA) is currently evaluating alternative treatment standards for radioactively contaminated high mercury (Hg) subcategory wastes, which do not require the removal of mercury from the waste. The Sulfur Polymer Stabilization/Solidification (SPSS) process developed at Brookhaven National Laboratory is one of several candidate technologies capable of successfully treating various Hg waste streams. To supplement previously supplied data on treatment of soils, EPA needed additional data concerning stabilization of high Hg subcategory waste sludges. To this end, a 5000 ppm sludge surrogate, containing approximately 50 wt% water, was successfully treated by pilot-scale SPSS processing. In two process runs, 85 and 95 wt% of water was recovered from the sludge during processing. At waste loadings of 46 wt% (30 wt% dry) sludge, the treated waste form had no detectable mercury (<10 ppb) in TCLP leachates. Data gathered from the demonstration of treatment of this sludge will provide the EPA with information to support revisions to current treatment requirements for high Hg subcategory wastes

  19. Allowable residual contamination levels: transuranic advanced disposal systems for defense waste

    International Nuclear Information System (INIS)

    Kennedy, W.E. Jr.; Napier, B.A.

    1982-01-01

    An evaluation of advanced disposal systems for defense transuranic (TRU) wastes is being conducted using the Allowable Residual Contamination Level (ARCL) method. The ARCL method is based on compliance with a radiation dose rate limit through a site-specific analysis of the potential for radiation exposure to individuals. For defense TRU wastes at the Hanford Site near Richland, Washington, various advanced disposal techniques are being studied to determine their potential for application. This paper presents a discussion of the results of the first stage of the TRU advanced disposal systems project

  20. Remote automatic control scheme for plasma arc cutting of contaminated waste

    International Nuclear Information System (INIS)

    Dudar, A.M.; Ward, C.R.; Kriikku, E.M.

    1993-01-01

    The Robotics Development Group at the Savannah River Technology Center has developed and implemented a scheme to perform automatic cutting of metallic contaminated waste. The scheme employs a plasma arc cutter in conjunction with a laser ranging sensor attached to a robotic manipulator called the Telerobot. A software algorithm using proportional control is then used to perturb the robot's trajectory in such a way as to regulate the plasma arc standoff and the robot's speed in order to achieve automatic plasma arc cuts. The scheme has been successfully tested on simulated waste materials and the results have been very favorable. This report details the development and testing of the scheme

  1. The use of supercritical carbon dioxide for contaminant removal from solid waste

    International Nuclear Information System (INIS)

    Adkins, C.L.J.; Russick, E.M.; Smith, H.M.; Olson, R.B.

    1994-01-01

    Supercritical carbon dioxide is being explored as a waste minimization technique for separating oils, greases and solvents from solid waste. The containments are dissolved into the supercritical fluid and precipitated out upon depressurization. The carbon dioxide solvent can then be recycled for continued use. Definitions of the temperature, pressure, flowrate and potential co-solvents are required to establish the optimum conditions for hazardous contaminant removal. Excellent extractive capability for common manufacturing oils, greases, and solvents has been observed in both supercritical and liquid carbon dioxide. Solubility measurements are being used to better understand the extraction process, and to determine if the minimum solubility required by federal regulations is met

  2. Eco-toxicity and metal contamination of paddy soil in an e-wastes recycling area

    International Nuclear Information System (INIS)

    Zhang Junhui; Hang Min

    2009-01-01

    Paddy soil samples taken from different sites in an old primitive electronic-waste (e-waste) processing region were examined for eco-toxicity and metal contamination. Using the environmental quality standard for soils (China, Grade II) as reference, soil samples of two sites were weakly contaminated with trace metal, but site G was heavily contaminated with Cd (6.37 mg kg -1 ), and weakly contaminated with Cu (256.36 mg kg -1 ) and Zn (209.85 mg kg -1 ). Zn appeared to be strongly bound in the residual fraction (72.24-77.86%), no matter the soil was metal contaminated or not. However, more than 9% Cd and 16% Cu was present in the non-residual fraction in the metal contaminated soils than in the uncontaminated soil, especially for site G and site F. Compared with that of the control soil, the micronucleus rates of site G and site F soil treatments increased by 2.7-fold and 1.7-fold, respectively. Low germination rates were observed in site C (50%) and site G (50%) soil extraction treated rice seeds. The shortest root length (0.2377 cm) was observed in site G soil treated groups, which is only 37.57% of that of the control soil treated groups. All of the micronucleus ratio of Vicia faba root cells, rice germination rate and root length after treatment of soil extraction indicate the eco-toxicity in site F and G soils although the three indexes are different in sensitivity to soil metal contamination.

  3. Foreign experience in alpha-contaminated waste disposal

    International Nuclear Information System (INIS)

    Lakey, P.

    1982-01-01

    The European presentations provided some useful comparisons with the situation int he United States regarding the transuranic (TRU) waste limit. First, in Europe, there appears to be a more moderate view on intrusion compared to the preoccupation in United States with this issue. Second, and superficially, in the United Kingdom and France, the working limit for near-surface disposal is greater than 10 nCi/g and more like 100 nCi/g. Looking beneath the superficial, however, the important difference is that their limits are working limits; they are not cast in bronze like the 10 nCi/g US value is not perceived to be. Europeans seem to have a more flexible and practical view of the issue and have reserved for its solution a rather large middle ground that appears to be lacking in the US position. For example, the United Kingdom is moving actively toward a version of greater confinement disposal or engineered disposal at a greater depth (with plutonium numbers like 10 4 nCi/g projected) and then moving on to the modified mine with limits like 10 5 nCi/g. From the French presentations, limits like 10 3 nCi/g were discussed. As we debate the TRU limit issue, what we seem to hear is an argument between the advocates of a generic limit of perhaps 100 nCi/g and the arguments for site-specific limits. This debate clouds perhaps the more basic issue of the need for a middle ground disposal approach between the extremes of a room trash limit and geologic disposal

  4. Design premises for canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    Werme, L.

    1998-09-01

    The purpose of this report is to establish the basic premises for designing canisters for the disposal of spent nuclear fuel, the requirements for canister characteristics, and the design criteria, and to present alternative canister designs that satisfy these premises. The point of departure for canister design has been that the canister must be able to be used for both BWR and PWR fuel

  5. Criticality safety study of Pu contaminated carbon waste stored in 100 L steel drums

    International Nuclear Information System (INIS)

    Anno, J.; Simonneau, M.

    1995-01-01

    The notion of the minimum critical areal density (D minca ) used to ensure the Criticality-Safety of poor solid waste is recalled with its deficiencies. D minca is assumed constant, independent of the fissile material concentration. This assumption is only true for unreflected mediums. Corrective factors are established. Furthermore, the usual norm of the Pu-H 2 O, which is 0.20 g/cm 2 , (concrete reflected) is greater than that for other mediums, such as Pu contaminated graphite waste (Pu-C), which is 0.036 g/cm 2 . D minca calculated on infinite slabs is confirmed by calculations on infinite planar multilayers arrays of 100 l cubical waste drums. Moreover, d minca increases linearly with the steel thickness of the drums' walls and goes up to 0.17 g/cm 2 for 0.105 cm of steel. The safety analysis on a real storage case takes into account the limited amount of Pu (100 g) and C (100 kg), the minimum thickness of 0.07 cm of drums' steel, their geometrical arrangement, the heterogeneity and size of contamination and the occurrence of neutronic poison (N and Cl) in the waste. Because of these parameters, the Keff are very less than 0.95 and the taken norm of 0.1 g/cm 2 for the Pu-C waste is fulfilled. Finally, it is demonstrated that the mixing of Pu-C waste drums and Pu-H 2 O wastes drums is allowed. (authors). 14 refs., 5 figs., 6 tabs

  6. Technical soaps - a possibility of decontaminating thorium-contaminated waste waters

    International Nuclear Information System (INIS)

    Drathen, H.; Erichsen, L. v.

    1977-01-01

    Thorium-contaminated waste waters showing a concentration of thorium higher than 10sup(-5) mol/l can be quantitatively decontaminated by adding soaps. Concentrations of impurity ions of both tap and sea waters have been taken into consideration. As there is no difference between soaps and soap mixtures concerning the quantity of precipitation rates, technical soaps are from the economic point of view best suited for decontaminating thorium-contaminated waste waters. Having a soap concentration of 200% of the stoichiometric amount of thorium and a concentration of impurity ions of 10sup(-2) mol/l, it is assumed that decontamination factors of more than 20 can be reached in one step. (orig.) [de

  7. Multi-purpose canisters as an alternative for storage, transportation, and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Hollaway, W.R.; Rozier, R.; Nitti, D.A.; Williams, J.R.

    1993-01-01

    A study was conducted to assess the feasibility of using multi-purpose canisters to handle spent nuclear fuel throughout the Civilian Radioactive Waste Management System. Multi-purpose canisters would be sealed, metallic containers maintaining multiple spent fuel assemblies in a dry, inert environment and overpacked separately and uniquely for the various system elements of storage, transportation, and disposal. Using five implementation scenarios, the multi-purpose canister was evaluated with regard to several measures of effectiveness, including number of handlings, radiation exposure, cost, schedule and licensing considerations, and public perception. Advantages and disadvantages of the multi-purpose canister were identified relative to the current reference system within each scenario, and the scenarios were compared to determine the most effective method of implementation

  8. Multiple-canister flow and transport code in 2-dimensional space. MCFT2D: user's manual

    International Nuclear Information System (INIS)

    Lim, Doo-Hyun

    2006-03-01

    A two-dimensional numerical code, MCFT2D (Multiple-Canister Flow and Transport code in 2-Dimensional space), has been developed for groundwater flow and radionuclide transport analyses in a water-saturated high-level radioactive waste (HLW) repository with multiple canisters. A multiple-canister configuration and a non-uniform flow field of the host rock are incorporated in the MCFT2D code. Effects of heterogeneous flow field of the host rock on migration of nuclides can be investigated using MCFT2D. The MCFT2D enables to take into account the various degrees of the dependency of canister configuration for nuclide migration in a water-saturated HLW repository, while the dependency was assumed to be either independent or perfectly dependent in previous studies. This report presents features of the MCFT2D code, numerical simulation using MCFT2D code, and graphical representation of the numerical results. (author)

  9. A review of groundwater contamination near municipal solid waste landfill sites in China.

    Science.gov (United States)

    Han, Zhiyong; Ma, Haining; Shi, Guozhong; He, Li; Wei, Luoyu; Shi, Qingqing

    2016-11-01

    Landfills are the most widely used method for municipal solid waste (MSW) disposal method in China. However, these facilities have caused serious groundwater contamination due to the leakage of leachate. This study, analyzed 32 scientific papers, a field survey and an environmental assessment report related to groundwater contamination caused by landfills in China. The groundwater quality in the vicinity of landfills was assessed as "very bad" by a comprehensive score (FI) of 7.85 by the Grading Method in China. Variety of pollutants consisting of 96 groundwater pollutants, 3 organic matter indicators, 2 visual pollutants and 6 aggregative pollutants had been detected in the various studies. Twenty-two kinds of pollutants were considered to be dominant. According to the Kruskal-Wallis test and the median test, groundwater contamination differed significantly between regions in China, but there were no significant differences between dry season and wet season measurements, except for some pollutants in a few landfill sites. Generally, the groundwater contamination appeared in the initial landfill stage after five years and peaked some years afterward. In this stage, the Nemerow Index (PI) of groundwater increased exponentially as landfill age increased at some sites, but afterwards decreased exponentially with increasing age at others. After 25years, the groundwater contamination was very low at selected landfills. The PI values of landfills decreased exponentially as the pollutant migration distance increased. Therefore, the groundwater contamination mainly appeared within 1000m of a landfill and most of serious groundwater contamination occurred within 200m. The results not only indicate that the groundwater contamination near MSW landfills should be a concern, but also are valuable to remediate the groundwater contamination near MSW landfills and to prevent the MSW landfill from secondary pollutions, especially for developing countries considering the similar

  10. Waste Area Grouping 2 Remedial Investigation Phase 1 Seep Task data report: Contaminant source area assessment

    International Nuclear Information System (INIS)

    Hicks, D.S.

    1996-03-01

    This report presents the findings of the Waste Area Grouping (WAG) 2, Phase 1 Remedial Investigation (RI) Seep Task efforts during 1993 and 1994 at Oak Ridge National Laboratory (ORNL). The results presented here follow results form the first year of sampling, 1992, which are contained in the Phase 1 RI report for WAG 2 (DOE 1995a). The WAG 2 Seep Task efforts focused on contaminants in seeps, tributaries, and main streams within the White Oak Creek (WOC) watershed. This report is designed primarily as a reference for contaminants and a resource for guiding remedial decisions. Additional in-depth assessments of the Seep Task data may provide clearer understandings of contaminant transport from the different source areas in the WOC watershed. WAG 2 consists of WOC and its tributaries downstream of the ORNL main plant area, White Oak Lake, the White Oak Creek Embayment of the Clinch River, and the associated flood plains and subsurface environment. The WOC watershed encompasses ORNL and associated WAGs. WAG 2 acts as an integrator for contaminant releases from the contaminated sites at ORNL and as the conduit transporting contaminants to the Clinch River. The main objectives of the Seep Task were to identify and characterize seeps, tributaries and source areas that are responsible for the contaminant releases to the main streams in WAG 2 and to quantify their input to the total contaminant release from the watershed at White Oak Dam (WOD). Efforts focused on 90 Sr, 3 H, and 137 Cs because these contaminants pose the greatest potential human health risk from water ingestion at WOD. Bimonthly sampling was conducted throughout the WOC watershed beginning in March 1993 and ending in August 1994. Samples were also collected for metals, anions, alkalinity, organics, and other radionuclides

  11. Multi-purpose canister system evaluation: A systems engineering approach

    International Nuclear Information System (INIS)

    1994-09-01

    This report summarizes Department of Energy (DOE) efforts to investigate various container systems for handling, transporting, storing, and disposing of spent nuclear fuel (SNF) assemblies in the Civilian Radioactive Waste Management System (CRWMS). The primary goal of DOE's investigations was to select a container technology that could handle the vast majority of commercial SNF at a reasonable cost, while ensuring the safety of the public and protecting the environment. Several alternative cask and canister concepts were evaluated for SNF assembly packaging to determine the most suitable concept. Of these alternatives, the multi-purpose canister (MPC) system was determined to be the most suitable. Based on the results of these evaluations, the decision was made to proceed with design and certification of the MPC system. A decision to fabricate and deploy MPCs will be made after further studies and preparation of an environmental impact statement

  12. Theoretical predictions for glass flow into an evacuated canister

    International Nuclear Information System (INIS)

    Routt, K.R.; Crow, K.R.

    1983-01-01

    Radioactive waste currently stored at the Savannah River Plant in liquid form is to be immobilized by incorporating it into a borosilicate glass. The glass melter for this process will consist of a refractory lined, steel vessel operated at a glass temperature of 1150 0 C. At the end of a two-year projected melter lifetime, the glass inside the melter is to be drained prior to disposition of the melter vessel. One proposed technique for accomplishing this drainage is by sucking the glass into an evacuated canister. The theoretical bases for design of an evacuated canister for draining a glass melter have been developed and tested. The theoretical equations governing transient and steady-state flow were substantiated with both a silicone glass simulant and molten glass

  13. Treatment of plutonium-contaminated solid waste: a review of handling systems

    International Nuclear Information System (INIS)

    Meredith, B.E.; Hardy, A.R.

    1985-02-01

    Handling techniques are reviewed to identify those suitable for adaptation for use in transporting large items of redundant plutonium contaminated plant and equipment to a remotely operated size reduction facility, moving them into the facility, presenting them to size reduction equipment and loading the processed waste into drums. It is concluded that an integrated system based on a combination of slatted conveyors, roller tables, air transporters and manipulators, merits further consideration. An appropriate experimental programme is outlined. (author)

  14. [Investigation of microbial contamination of the air and equipment of a biological waste water purification station].

    Science.gov (United States)

    Alikbaeva, L A; Figurovskiĭ, A P; Vasil'ev, O D; Ermolaev-Makovskiĭ, M A; Merkur'eva, M A

    2010-01-01

    The paper describes the results of a study of ambient air microbiological pollution in the working premises and equipment surfaces in the main shops of the biological waste water purification station of a cardboard-polygraphic plant. The findings suggest that there is high microbial contamination of the working environment, which should be born in mind on developing measures to optimize working conditions and on studying morbidity rates among the workers.

  15. Hanford Tank 241-C-106: Residual Waste Contaminant Release Model and Supporting Data

    International Nuclear Information System (INIS)

    Deutsch, William J.; Krupka, Kenneth M.; Lindberg, Michael J.; Cantrell, Kirk J.; Brown, Christopher F.; Schaef, Herbert T.

    2005-01-01

    CH2M HILL is producing risk/performance assessments to support the closure of single-shell tanks at the DOE's Hanford Site. As part of this effort, staff at PNNL were asked to develop release models for contaminants of concern that are present in residual sludge remaining in tank 241-C-106 (C-106) after final retrieval of waste from the tank. This report provides the information developed by PNNL

  16. Manufacture and inspection of metal containers for the storage of waste contaminated with caesium-137

    International Nuclear Information System (INIS)

    Brito, M.J.D.; Moreira, M.S.

    1998-01-01

    Several stages are described of the design and manufacture of a prototype unit and 15 metal cylindrical containers for the storage of contaminated waste generated by the radiological accident of Goiania in 1987. The tasks involved technicians from the Nuclear Technology Development Centre (CDTN) of Belo Horizonte and the co-authors, who conducted inspections in the period between 10 March 1993 and 18 May 1993 at the foundry in Goiania. (author)

  17. Spatial distribution of heavy metal contamination in soils near a primitive e-waste recycling site.

    Science.gov (United States)

    Quan, Sheng-Xiang; Yan, Bo; Yang, Fan; Li, Ning; Xiao, Xian-Ming; Fu, Jia-Mo

    2015-01-01

    The total concentrations of 12 heavy metals in surface soils (SS, 0-20 cm), middle soils (MS, 30-50 cm) and deep soils (DS, 60-80 cm) from an acid-leaching area, a deserted paddy field and a deserted area of Guiyu were measured. The results showed that the acid-leaching area was heavily contaminated with heavy metals, especially in SS. The mean concentrations of Ni, Cu, Zn, Cd, Sn, Sb and Pb in SS from the acid-leaching area were 278.4, 684.1, 572.8, 1.36, 3,472, 1,706 and 222.8 mg/kg, respectively. Heavy metal pollution in the deserted paddy field was mainly concentrated in SS and MS. The average values of Sb in SS and MS from the deserted paddy field were 16.3 and 20.2 mg/kg, respectively. However, heavy metal contamination of the deserted area was principally found in the DS. Extremely high concentrations of heavy metals were also observed at some special research sites, further confirming that the level of heavy metal pollution was very serious. The geoaccumulation index (Igeo) values revealed that the acid-leaching area was severely polluted with heavy metals in the order of Sb > Sn > Cu > Cd > Ni > Zn > Pb, while deserted paddy field was contaminated predominately by metals in the order of Sb > Sn > Cu. It was obvious that the concentrations of some uncommon contaminants, such as Sb and Sn, were higher than principal contaminants, such as Ni, Cu, Zn and Pb, suggesting that particular attention should be directed to Sn and Sb contamination in the future research of heavy metals in soils from e-waste-processing areas. Correlation analysis suggested that Li and Be in soils from the acid-leaching area and its surrounding environment might have originated from other industrial activities and from batteries, whereas Ni, Cu, Zn, Cd, Pb, Sn and Sb contamination was most likely caused by uncontrolled electronic waste (e-waste) processing. These results indicate the significant need for optimisation of e-waste-dismantling technologies and remediation of polluted soil

  18. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    Energy Technology Data Exchange (ETDEWEB)

    Wishau, R.

    1998-05-01

    Molten salt oxidation (MSO) is proposed as a {sup 238}Pu waste treatment technology that should be developed for volume reduction and recovery of {sup 238}Pu and as an alternative to the transport and permanent disposal of {sup 238}Pu waste to the WIPP repository. In MSO technology, molten sodium carbonate salt at 800--900 C in a reaction vessel acts as a reaction media for wastes. The waste material is destroyed when injected into the molten salt, creating harmless carbon dioxide and steam and a small amount of ash in the spent salt. The spent salt can be treated using aqueous separation methods to reuse the salt and to recover 99.9% of the precious {sup 238}Pu that was in the waste. Tests of MSO technology have shown that the volume of combustible TRU waste can be reduced by a factor of at least twenty. Using this factor the present inventory of 574 TRU drums of {sup 238}Pu contaminated wastes is reduced to 30 drums. Further {sup 238}Pu waste costs of $22 million are avoided from not having to repackage 312 of the 574 drums to a drum total of more than 4,600 drums. MSO combined with aqueous processing of salts will recover approximately 1.7 kilograms of precious {sup 238}Pu valued at 4 million dollars (at $2,500/gram). Thus, installation and use of MSO technology at LANL will result in significant cost savings compared to present plans to transport and dispose {sup 238}Pu TRU waste to the WIPP site. Using a total net present value cost for the MSO project as $4.09 million over a five-year lifetime, the project can pay for itself after either recovery of 1.6 kg of Pu or through volume reduction of 818 drums or a combination of the two. These savings show a positive return on investment.

  19. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    International Nuclear Information System (INIS)

    Wishau, R.

    1998-05-01

    Molten salt oxidation (MSO) is proposed as a 238 Pu waste treatment technology that should be developed for volume reduction and recovery of 238 Pu and as an alternative to the transport and permanent disposal of 238 Pu waste to the WIPP repository. In MSO technology, molten sodium carbonate salt at 800--900 C in a reaction vessel acts as a reaction media for wastes. The waste material is destroyed when injected into the molten salt, creating harmless carbon dioxide and steam and a small amount of ash in the spent salt. The spent salt can be treated using aqueous separation methods to reuse the salt and to recover 99.9% of the precious 238 Pu that was in the waste. Tests of MSO technology have shown that the volume of combustible TRU waste can be reduced by a factor of at least twenty. Using this factor the present inventory of 574 TRU drums of 238 Pu contaminated wastes is reduced to 30 drums. Further 238 Pu waste costs of $22 million are avoided from not having to repackage 312 of the 574 drums to a drum total of more than 4,600 drums. MSO combined with aqueous processing of salts will recover approximately 1.7 kilograms of precious 238 Pu valued at 4 million dollars (at $2,500/gram). Thus, installation and use of MSO technology at LANL will result in significant cost savings compared to present plans to transport and dispose 238 Pu TRU waste to the WIPP site. Using a total net present value cost for the MSO project as $4.09 million over a five-year lifetime, the project can pay for itself after either recovery of 1.6 kg of Pu or through volume reduction of 818 drums or a combination of the two. These savings show a positive return on investment

  20. Initial site characterization and evaluation of radionuclide contaminated soil waste burial grounds

    International Nuclear Information System (INIS)

    Phillips, S.J.; Reisenauer, A.E.; Rickard, W.H.; Sandness, G.A.

    1977-02-01

    A survey of historical records and literature containing information on the contents of 300 Area and North Burial Grounds was completed. Existing records of radioactive waste location, type, and quantity within each burial ground facility were obtained and distributed to cooperating investigators. A study was then initiated to evaluate geophysical exploration techniques for mapping buried waste materials, waste containers, and trench boundaries. Results indicate that a combination of ground penetrating radar, magnetometer, metal detector, and acoustic measurements will be effective but will require further study, hardware development, and field testing. Drilling techniques for recovering radionuclide-contaminated materials and sediment cores were developed and tested. Laboratory sediment characterization and fluid transport and monitoring analyses were begun by installation of in situ transducers at the 300 North Burial Ground site. Biological transport mechanisms that control radionuclide movement at contaminated sites were also studied. Flora and fauna presently inhabiting specific burial ground areas were identified and analyzed. Future monitoring of specific mammal populations will permit determination of dose rate and pathways of contaminated materials contained in and adjacent to burial ground sites

  1. West Siberian basin hydrogeology - regional framework for contaminant migration from injected wastes

    International Nuclear Information System (INIS)

    Foley, M.G.

    1994-05-01

    Nuclear fuel cycle activities of the former Soviet Union (FSU) have resulted in massive contamination of the environment in western Siberia. We are developing three-dimensional numerical models of the hydrogeology and potential contaminant migration in the West Siberian Basin. Our long-term goal at Pacific Northwest Laboratory is to help determine future environmental and human impacts given the releases that have occurred to date and the current waste management practices. In FY 1993, our objectives were to (1) refine and implement the hydrogeologic conceptual models of the regional hydrogeology of western Siberia developed in FY 1992 and develop the detailed, spatially registered digital geologic and hydrologic databases to test them, (2) calibrate the computer implementation of the conceptual models developed in FY 1992, and (3) develop general geologic and hydrologic information and preliminary hydrogeologic conceptual models relevant to the more detailed models of contaminated site hydrogeology. Calibration studies of the regional hydrogeologic computer model suggest that most precipitation entering the ground-water system moves in the near-surface part of the system and discharges to surface waters relatively near its point of infiltration. This means that wastes discharged to the surface and near-surface may not be isolated as well as previously thought, since the wastes may be carried to the surface by gradually rising ground waters

  2. Variability in physical contamination assessment of source segregated biodegradable municipal waste derived composts.

    Science.gov (United States)

    Echavarri-Bravo, Virginia; Thygesen, Helene H; Aspray, Thomas J

    2017-01-01

    Physical contaminants (glass, metal, plastic and 'other') and stones were isolated and categorised from three finished commercial composts derived from source segregated biodegradable municipal waste (BMW). A subset of the identified physical contaminant fragments were subsequently reintroduced into the cleaned compost samples and sent to three commercial laboratories for testing in an inter-laboratory trial using the current PAS100:2011 method (AfOR MT PC&S). The trial showed that the 'other' category caused difficulty for all three laboratories with under reporting, particularly of the most common 'other' contaminants (paper and cardboard) and, over-reporting of non-man-made fragments. One laboratory underreported metal contaminant fragments (spiked as silver foil) in three samples. Glass, plastic and stones were variably underreported due to miss-classification or over reported due to contamination with compost (organic) fragments. The results are discussed in the context of global physical contaminant test methods and compost quality assurance schemes. Copyright © 2016 Elsevier Ltd. All rights reserved.

  3. Nuclear Wastes in the Arctic: An Analysis of Arctic and Other Regional Impacts from Soviet Nuclear Contamination

    National Research Council Canada - National Science Library

    1995-01-01

    ..., and from radioactive releases from both past and future nuclear activities in the region. The report presents what is known and unknown about this waste and contamination and how it may affect public health...

  4. Canisters for spent-fuel disposal: Design measures against localized corrosion

    International Nuclear Information System (INIS)

    Werme, L.O.; Oversby, V.M.

    2000-01-01

    Common to all high-level-waste disposal concepts is the encapsulation of the waste into metal canisters. The purpose of this waste canister is to isolate the radioactive waste from contact with its surroundings for a desired time period. The design service life ranges from hundreds to thousands of years depending on the disposal concept. After the isolation has been breached, other barriers in the disposal system will delay and attenuate the radioactive releases to acceptable levels. In a deep geologic repository, the waste package will be exposed to chemical attack and, depending on the type of repository, to mechanical stresses. Each of these factors will by itself or in combination inevitably lead to loss of confinement some time in the future. In the design of the Swedish waste canister, the corrosion resistance is provided by an outer shell of pure copper while an insert supplies the mechanical strength cast nodular iron. The close fit between the insert and the copper results in very small tensile stresses in the copper over very limited areas once the repository has been saturated. Measurements of stress corrosion crack growth show that annealed copper cannot maintain sufficiently high stress intensity factors for cracks to grow. For annealed copper, the stress intensity factor was limited to 25 MPa·m 1/2 because of extensive plastic deformation. For cold-worked copper, no crack growth could be observed for stress intensity factors 1/2 . Through the choices of canister material, canister, and repository design, and considering the expected chemical conditions, the risks for localized corrosion can be lowered to an acceptable level, if not eliminated altogether, and the releases from prematurely failed canisters can be kept well within acceptable dose levels

  5. Levels for the specific activity at disposing low-level contaminated municipal wastes

    International Nuclear Information System (INIS)

    Poschner, J.; Schaller, G.

    1995-01-01

    Using radioecological models, nuclide specific values were calculated for the specific activity of low contaminated radioactive waste, which is disposed in conventional waste deposits or burned in incineration plants. The calculation of these values is based on a limit of 10 μSv effective dose in one year, i.e. effective dose possibly resulting from waste disposal or burning should not exceed a 'de-minimis'-value of some 10 μSv per year. The applied radioecological models describe exposure of the public by direct radiation, inhalation and ingestion for the operational period of a deposit or an incineration plant, but also cover post-operational scenarios, collecting and sorting of waste and road accidents of the waste-truck. Referring to the dose limit of 10 μSv/a, a value for the specific activity of waste was calculated for each scenario and each radionuclide considered. The smallest of these values for a radionuclide, the 'basic value' was rounded to a 'reference value'. For about 600 radionuclides reference values were derived. About 90% of the reference values are ranging between 1 and 1 000 Bq/g. For about 90% of the radionuclides direct radiation or inhalation at the deposit proved to be the critical path of exposure. (orig.) [de

  6. Phytoremediation trials on metal- and arsenic-contaminated pyrite wastes (Torviscosa, Italy)

    Energy Technology Data Exchange (ETDEWEB)

    Vamerali, Teofilo [Department of Environmental Sciences, University of Parma, Viale G.P. Usberti 11/A, 43100 Parma (Italy)], E-mail: teofilo.vamerali@unipd.it; Bandiera, Marianna; Coletto, Lucia; Zanetti, Federica [Department of Environmental Agronomy and Crop Sciences, University of Padova, Viale dell' Universita 16, 35020 Legnaro - Padova (Italy); Dickinson, Nicholas M. [Faculty of Science, Liverpool John Moores University, Byrom Street, Liverpool, L3 3AF (United Kingdom); Mosca, Giuliano [Department of Environmental Agronomy and Crop Sciences, University of Padova, Viale dell' Universita 16, 35020 Legnaro - Padova (Italy)

    2009-03-15

    At a site in Udine, Italy, a 0.7 m layer of As, Co, Cu, Pb and Zn contaminated wastes derived from mineral roasting for sulphur extraction had been covered with an unpolluted 0.15 m layer of gravelly soil. This study investigates whether woody biomass phytoremediation is a realistic management option. Comparing ploughing and subsoiling (0.35 m depth), the growth of Populus and Salix and trace element uptake were investigated in both pot and field trials. Species differences were marginal and species selection was not critical. Impaired above-ground productivity and low translocation of trace elements showed that bioavailable contaminant stripping was not feasible. The most significant finding was of coarse and fine roots proliferation in surface layers that provided a significant sink for trace elements. We conclude that phytostabilisation and effective immobilisation of metals and As could be achieved at the site by soil amelioration combined with woody species establishment. Confidence to achieve a long-term and sustainable remediation requires a more complete quantification of root dynamics and a better understanding of rhizosphere processes. - In As- and metal-contaminated pyrite wastes, contaminant stripping is not feasible, and root foraging and quantification of root dynamics holds the key to stabilisation in woody species.

  7. Phytoremediation trials on metal- and arsenic-contaminated pyrite wastes (Torviscosa, Italy)

    International Nuclear Information System (INIS)

    Vamerali, Teofilo; Bandiera, Marianna; Coletto, Lucia; Zanetti, Federica; Dickinson, Nicholas M.; Mosca, Giuliano

    2009-01-01

    At a site in Udine, Italy, a 0.7 m layer of As, Co, Cu, Pb and Zn contaminated wastes derived from mineral roasting for sulphur extraction had been covered with an unpolluted 0.15 m layer of gravelly soil. This study investigates whether woody biomass phytoremediation is a realistic management option. Comparing ploughing and subsoiling (0.35 m depth), the growth of Populus and Salix and trace element uptake were investigated in both pot and field trials. Species differences were marginal and species selection was not critical. Impaired above-ground productivity and low translocation of trace elements showed that bioavailable contaminant stripping was not feasible. The most significant finding was of coarse and fine roots proliferation in surface layers that provided a significant sink for trace elements. We conclude that phytostabilisation and effective immobilisation of metals and As could be achieved at the site by soil amelioration combined with woody species establishment. Confidence to achieve a long-term and sustainable remediation requires a more complete quantification of root dynamics and a better understanding of rhizosphere processes. - In As- and metal-contaminated pyrite wastes, contaminant stripping is not feasible, and root foraging and quantification of root dynamics holds the key to stabilisation in woody species

  8. Bioremediation of oil-contaminated soil using Candida catenulata and food waste.

    Science.gov (United States)

    Joo, Hung-Soo; Ndegwa, Pius M; Shoda, Makoto; Phae, Chae-Gun

    2008-12-01

    Even though petroleum-degrading microorganisms are widely distributed in soil and water, they may not be present in sufficient numbers to achieve contaminant remediation. In such cases, it may be useful to inoculate the polluted area with highly effective petroleum-degrading microbial strains to augment the exiting ones. In order to identify a microbial strain for bioaugmentation of oil-contaminated soil, we isolated a microbial strain with high emulsification and petroleum hydrocarbon degradation efficiency of diesel fuel in culture. The efficacy of the isolated microbial strain, identified as Candida catenulata CM1, was further evaluated during composting of a mixture containing 23% food waste and 77% diesel-contaminated soil including 2% (w/w) diesel. After 13 days of composting, 84% of the initial petroleum hydrocarbon was degraded in composting mixes containing a powdered form of CM1 (CM1-solid), compared with 48% of removal ratio in control reactor without inoculum. This finding suggests that CM1 is a viable microbial strain for bioremediation of oil-contaminated soil with food waste through composting processes.

  9. Estimation of doses from radioactively contaminated disaster wastes reused for pavements

    International Nuclear Information System (INIS)

    Sawaguchi, Takuma; Takeda, Seiji; Kimura, Hideo; Tanaka, Tadao

    2015-01-01

    It is desirable that the disaster wastes contaminated by radioactive cesium after the severe accident at the Fukushima Nuclear Plant are reused as much as possible in order to minimize the quantity to be disposed of. Ministry of the Environment showed the policy that the wastes containing cesium of higher concentration than the clearance level (100 Bq/kg) were reusable as materials of construction such as subbase course materials of pavements under controlled condition with measures to lower exposure doses. In this study, in order to provide technical information for making a guideline on the use of contaminated concrete materials recycled from disaster wastes as pavement, doses for workers and the public were estimated, and the reusable concentration of radioactive cesium in the wastes was evaluated. It was shown that the external exposure of the public (children) residing near the completed pavement gave the minimum radiocesium concentration in order to comply with the dose criteria. The recycled concrete materials whose average concentration of cesium lower than 2,700 Bq/kg can be used as the subbase course materials of pavements. (author)

  10. Preliminary studies on trace element contamination in dumping sites of municipal wastes in India and Vietnam

    Science.gov (United States)

    Agusa, T.; Kunito, T.; Nakashima, E.; Minh, T. B.; Tanabe, S.; Subramanian, A.; Viet, P. H.

    2003-05-01

    The disposal of wastes in dumping sites has increasingly caused concem about adverse health effects on the populations living nearby. However, no investigation has been conducted yet on contamination in dumping sites of municipal wastes in Asian developing countries. In this study, concentrations of 11 trace elements (V, Cr, Mn, Co, Cu, Zn, Mo, Ag, Cd, Sb and Pb) were detennined in scalp hair from the population living nearby and in soil from dumping sites and control sites of India and Vietnam. Soil samples in dumping site in India showed significantly higher concentrations of some trace elements than soils in control site, whereas this trend was not notable in Vietnam. This is probably due to the fact that the wastes were covered with the soil in the dumping site of Vietnam. Cadmium concentrations in some hair samples of people living near dumping site in India and Vietnam exceeded the level associated with learning disorder in children. Levels of most of the trace elements in hair were significantly higher in dumping site than those in control site in India and Vietnam, suggesting direct or indirect exposure to those elements from dumping wastes. To our knowledge, this is the first study of trace element contamination in dumping sites in India and Vietnam.

  11. Closure Report for Corrective Action Unit 143: Area 25 Contaminated Waste Dumps, Nevada Test Site, Nevada

    International Nuclear Information System (INIS)

    Tobiason, D. S.

    2002-01-01

    This Closure Report (CR) has been prepared for the Area 25 Contaminated Waste Dumps (CWD), Corrective Action Unit (CAU) 143 in accordance with the Federal Facility Agreement and Consent Order [FFACO] (FFACO, 1996) and the Nevada Division of Environmental Protection (NDEP)-approved Corrective Action Plan (CAP) for CAU 143: Area 25, Contaminated Waste Dumps, Nevada Test Site, Nevada. CAU 143 consists of two Corrective Action Sites (CASs): 25-23-09 CWD No.1, and 25-23-03 CWD No.2. The Area 25 CWDs are historic disposal units within the Area 25 Reactor Maintenance, Assembly, and Disassembly (R-MAD), and Engine Maintenance, Assembly, and Disassembly (E-MAD) compounds located on the Nevada Test Site (NTS). The R-MAD and E-MAD facilities originally supported a portion of the Nuclear Rocket Development Station in Area 25 of the NTS. CWD No.1 CAS 25-23-09 received solid radioactive waste from the R-MAD Compound (East Trestle and West Trench Berms) and 25-23-03 CWD No.2 received solid radioactive waste from the E-MAD Compound (E-MAD Trench)

  12. Processing results of 1800 gallons of mercury and radioactively contaminated mixed waste rinse solution

    International Nuclear Information System (INIS)

    Thiesen, B.P.

    1993-01-01

    Mercury-contaminated rinse solution was successfully treated at the Idaho National Engineering Laboratory. This waste was generated during the decontamination of the Heat Transfer Reactor Experiment 3 reactor shield tank. Approximately 6.8 m 3 (1,800 pi) of waste was generated and placed into 33 drums. Each drum contained precipitated sludge material ranging from 2--5 cm in depth, with the average depth of about 6 cm. The pH of each drum varied from 3--11. The bulk liquid waste had a mercury level of 7.0 mg/l, which exceeded the Resource Conservation and Recovery Act limit of 0.2 mg/l. The average liquid bulk radioactivity was about 2.1 pCi/mL while the average sludge contamination was about 13,800 pCi/g. Treatment of the waste required separation of the liquid from the sludge, filtration, pH adjustment, and ion exchange. The resulting solution after treatment had mercury levels at 0.0186 mg/l and radioactivity of 0.282 pCi/ml

  13. Food contamination by PCBs and waste disposal crisis: Evidence from goat milk in Campania (Italy).

    Science.gov (United States)

    Ferrante, M C; Fusco, G; Monnolo, A; Saggiomo, F; Guccione, J; Mercogliano, R; Clausi, M T

    2017-11-01

    The study aims at investigating whether, and if so, to what extent the strong presence of urban and industrial waste in a territory may cause PCB contamination in goat milk produced therein. We compared PCB concentrations in goat milk from three different locations in the Campania region (Italy). One of the three locations, together with its surrounding area, has long suffered from illegal waste disposal and burning mainly by the so-called Ecomafia. The other locations, not involved in these illegal activities, allowed us to create a control group of goats with characteristics very similar to those of main interest. In milk from the waste contaminated area we identified high PCB concentrations (six indicator PCBs amounted to 170 ng g -1 on lipid weight, on average), whereas there was an almost total absence of such pollutants in milk from the control group. Concentrations of the six indicator PCBs were above the current European maximum residue limit fixed by the EU. At the same time, we found a lower average value of lipid content and a negative relationship between lipid content and PCB concentrations. Evidence indicates the potential health risk for consumers living in areas involved in illegal dumping of waste. Copyright © 2017 Elsevier Ltd. All rights reserved.

  14. Caffeine and pharmaceuticals as indicators of waste water contamination in wells

    Science.gov (United States)

    Seiler, R.L.; Zaugg, S.D.; Thomas, J.M.; Howcroft, D.L.

    1999-01-01

    The presence of caffeine or human pharmaceuticals in ground water with elevated nitrate concentrations can provide a clear, unambiguous indication that domestic waste water is a source of some of the nitrate. Water from domestic, public supply, and monitoring wells in three communities near Reno, Nevada, was sampled to test if caffeine or pharmaceuticals are common, persistent, and mobile enough in the environment that they can be detected in nitrate-contaminated ground water and, thus, can be useful indicators of recharge from domestic waste water. Results of this study indicate that these compounds can be used as indicators of recharge from domestic waste water, although their usefulness is limited because caffeine is apparently nonconservative and the presence of prescription pharmaceuticals is unpredictable. The absence of caffeine or pharmaceuticals in ground water with elevated nitrate concentrations does not demonstrate that the aquifer is free of waste water contamination. Caffeine was detected in ground water samples at concentrations up to 0.23 ??g/L. The human pharmaceuticals chlorpropamide, phensuximide, and carbamazepine also were detected in some samples.

  15. Barrier capacity of weathered coal mining wastes with respect to heavy metal and organic contaminants

    International Nuclear Information System (INIS)

    Twardowska, I.; Jarosinska, B.

    1992-01-01

    Some types of weathered, buffered coal mining wastes (CMW), being essentially heterogenous and complex mineralogical system of developed surface area, under certain conditions could be widely applicable for binding a variety of contaminants both inorganic in cationic or anionic form, and organic compounds. The experiments reported earlier, showed excellent Cr(VI)-binding capacity of CMW. In this paper, experiments on simultaneous removal of heavy metals Cr t , Cu 2+ , Zn 2+ and Cd 2+ from highly (pH 2.5) and mildly acidic solutions (pH 4.0), as well as of organic compounds and color reduction in leachate from solid industrial waste dump (foundry wastes) will be presented

  16. Behavior of radioactive cesium during incineration of radioactively contaminated wastes from decontamination activities in Fukushima.

    Science.gov (United States)

    Fujiwara, Hiroshi; Kuramochi, Hidetoshi; Nomura, Kazutaka; Maeseto, Tomoharu; Osako, Masahiro

    2017-11-01

    Large volumes of decontamination wastes (DW) generated by off-site decontamination activities in Fukushima Prefecture have been incinerated since 2015. The behavior of radioactive cesium during incineration of DW was investigated at a working incineration plant. The incineration discharged bottom ash (BA) and fly ash (FA) with similar levels of radiocesium, and the leachability of the radiocesium from both types of ash was very low (incineration of contaminated municipal solid waste (CMSW) reported in earlier studies. The source of radiocesium in DW-FA is chiefly small particles derived from DW and DW-BA blown into the flue gas, not the deposition of gaseous synthesized radiocesium compounds on the surfaces of ash particles in the flue gas as observed in CMSW incineration. This source difference causes the behavior of radiocesium during waste incineration to differ between DW and CMSW. Copyright © 2017 Elsevier Ltd. All rights reserved.

  17. A convenient method for estimating the contaminated zone of a subsurface aquifer resulting from radioactive waste disposal into ground

    International Nuclear Information System (INIS)

    Fukui, Masami; Katsurayama, Kousuke; Uchida, Shigeo.

    1981-01-01

    Studies were conducted to estimate the contamination spread resulting from the radioactive waste disposal into a subsurface aquifer. A general equation, expressing the contaminated zone as a function of radioactive decay, the physical and chemical parameters of soil is presented. A distribution coefficient was also formulated which can be used to judge the suitability of a site for waste disposal. Moreover, a method for predicting contaminant concentration in groundwater at a site boundary is suggested for a heterogeneous media where the subsurface aquifer has different values of porosity, density, flow velocity, distribution coefficient and so on. A general equation was also developed to predict the distribution of radionuclides resulting from the disposal of a solid waste material. The distributions of contamination was evaluated for 90 Sr and 239 Pu which obey a linear adsorption model and a first order kinetics respectively. These equations appear to have practical utility for easily estimating groundwater contamination. (author)

  18. Municipal waste management and groundwater contamination processes in Córdoba Province, Argentina

    Directory of Open Access Journals (Sweden)

    Daniel Emilio Martínez

    2010-12-01

    Full Text Available In Coronel Moldes, Argentina, waste management practices consist in municipal waste being tipped directly onto an area of sand dunes at the municipal waste disposal site (MWDS. Moreover, untreated liquid waste from septic tanks and latrines from urban areas are discharged in the same place. This co-disposal waste management is very common in many regions of Argentina and its impact on the groundwater of Coronel Moldes has not been evaluated. The study area is located in the vicinity of a MWDS in a flatlands environment that is typical of Argentina. The main objective of this study was to evaluate the impacts on groundwater quality of current waste management practices in order to consider the requirement for new guidelines for sustainable groundwater management. Three groundwater monitoring wells were installed up-, across- and down-gradient of the MWDS. The principal aquifer is formed by sandy silt sediments (loess. Groundwater levels in the area of the MWDS are between 5.6 m and 7.8 m. The Vulnerability index indicates that groundwater in this area has a high vulnerability. Groundwater in the vicinity of the MWDS shows elevated electrical conductivity, high concentrations of Cl-, Na+, and HCO3- ions, COD, BOD5 and aerobic bacteria and less dissolved oxygen than the background values indicating the presence of organic matter. Municipal waste management represents a significant omission in current groundwater protection policy at Coronel Moldes. Strict supervision of solid and liquid municipal waste disposal needs to be instigated in order to ensure that the groundwater remains free of contamination and to allow a sustainable environmental management.

  19. Application of eggshell waste for the immobilization of cadmium and lead in a contaminated soil.

    Science.gov (United States)

    Ok, Yong Sik; Lee, Sang Soo; Jeon, Weon-Tai; Oh, Sang-Eun; Usman, Adel R A; Moon, Deok Hyun

    2011-01-01

    Liming materials have been used to immobilize heavy metals in contaminated soils. However, no studies have evaluated the use of eggshell waste as a source of calcium carbonate (CaCO₃) to immobilize both cadmium (Cd) and lead (Pb) in soils. This study was conducted to evaluate the effectiveness of eggshell waste on the immobilization of Cd and Pb and to determine the metal availability following various single extraction techniques. Incubation experiments were conducted by mixing 0-5% powdered eggshell waste and curing the soil (1,246 mg Pb kg⁻¹ soil and 17 mg Cd kg⁻¹ soil) for 30 days. Five extractants, 0.01 M calcium chloride (CaCl₂), 1 M CaCl₂, 0.1 M hydrochloric acid (HCl), 0.43 M acetic acid (CH₃COOH), and 0.05 M ethylendiaminetetraacetic acid (EDTA), were used to determine the extractability of Cd and Pb following treatments with CaCO₃ and eggshell waste. Generally, the extractability of Cd and Pb in the soils decreased in response to treatments with CaCO₃ and eggshell waste, regardless of extractant. Using CaCl₂ extraction, the lowest Cd concentration was achieved upon both CaCO₃ and eggshell waste treatments, while the lowest Pb concentration was observed using HCl extraction. The highest amount of immobilized Cd and Pb was extracted by CH₃COOH or EDTA in soils treated with CaCO₃ and eggshell waste, indicating that remobilization of Cd and Pb may occur under acidic conditions. Based on the findings obtained, eggshell waste can be used as an alternative to CaCO₃ for the immobilization of heavy metals in soils.

  20. Inspection of disposal canisters components

    International Nuclear Information System (INIS)

    Pitkaenen, J.

    2013-12-01

    This report presents the inspection techniques of disposal canister components. Manufacturing methods and a description of the defects related to different manufacturing methods are described briefly. The defect types form a basis for the design of non-destructive testing because the defect types, which occur in the inspected components, affect to choice of inspection methods. The canister components are to nodular cast iron insert, steel lid, lid screw, metal gasket, copper tube with integrated or separate bottom, and copper lid. The inspection of copper material is challenging due to the anisotropic properties of the material and local changes in the grain size of the copper material. The cast iron insert has some acoustical material property variation (attenuation, velocity changes, scattering properties), which make the ultrasonic inspection demanding from calibration point of view. Mainly three different methods are used for inspection. Ultrasonic testing technique is used for inspection of volume, eddy current technique, for copper components only, and visual testing technique are used for inspection of the surface and near surface area

  1. Mercury contamination and potential impacts from municipal waste incinerator on Samui Island, Thailand.

    Science.gov (United States)

    Muenhor, Dudsadee; Satayavivad, Jutamaad; Limpaseni, Wongpun; Parkpian, Preeda; Delaune, R D; Gambrell, R P; Jugsujinda, Aroon

    2009-03-01

    In recent years, mercury (Hg) pollution generated by municipal waste incinerators (MWIs) has become the subject of serious public concern. On Samui Island, Thailand, a large-scale municipal waste incinerator has been in operation for over 7 years with a capacity of 140 tons/day for meeting the growing demand for municipal waste disposal. This research assessed Hg contamination in environmental matrices adjacent to the waste incinerating plant. Total Hg concentrations were determined in municipal solid waste, soil and sediment within a distance of 100 m to 5 km from the incinerator operation in both wet and dry seasons. Hg analyses conducted in municipal solid waste showed low levels of Hg ranging between 0.15-0.56 mg/kg. The low level was due to the type of waste incinerator. Waste such as electrical appliances, motors and spare parts, rubber tires and hospital wastes are not allowed to feed into the plant. As a result, low Hg levels were also found in fly and bottom ashes (0.1-0.4 mg/kg and Stack concentration of Hg were less than 0.4 microg/Nm(3). Since Hg emissions were at low concentrations, Hg in soil from atmospheric fallout near this incinerator including uptake by local weeds were very low ranging from non detectable to 399 micro g/kg. However, low but elevated levels of Hg (76-275 micro g/kg) were observed in surface soil and deeper layers (0-40 cm) in the predominant downwind direction of incinerator over a distance of between 0.5-5 km. Soil Hg concentrations measured from a reference/background track opposite of the prevailing wind direction were lower ranging between 7-46 micro g/kg. Nevertheless, the trend of Hg build up in soil was clearly seen in the wet season only, suggesting that wet deposition process is a major Hg pollution source. Hg concentrations in the sea bottom sediment collected next to the last station track was small with values between 35-67 micro g/kg. Based upon the overall findings, in terms of current potential environmental risk

  2. Demonstration of New Technologies Required for the Treatment of Mixed Waste Contaminated with {ge}260 ppm Mercury

    Energy Technology Data Exchange (ETDEWEB)

    Morris, M.I.

    2002-02-06

    The Resource Conservation and Recovery Act (RCRA) defines several categories of mercury wastes, each of which has a defined technology or concentration-based treatment standard, or universal treatment standard (UTS). RCRA defines mercury hazardous wastes as any waste that has a TCLP value for mercury of 0.2 mg/L or greater. Three of these categories, all nonwastewaters, fall within the scope of this report on new technologies to treat mercury-contaminated wastes: wastes as elemental mercury; hazardous wastes with less than 260 mg/kg [parts per million (ppm)] mercury; and hazardous wastes with 260 ppm or more of mercury. While this report deals specifically with the last category--hazardous wastes with 260 ppm or more of mercury--the other two categories will be discussed briefly so that the full range of mercury treatment challenges can be understood. The treatment methods for these three categories are as follows: Waste as elemental mercury--RCRA identifies amalgamation (AMLGM) as the treatment standard for radioactive elemental mercury. However, radioactive mercury condensates from retorting (RMERC) processes also require amalgamation. In addition, incineration (IMERC) and RMERC processes that produce residues with >260 ppm of radioactive mercury contamination and that fail the RCRA toxicity characteristic leaching procedure (TCLP) limit for mercury (0.20 mg/L) require RMERC, followed by AMLGM of the condensate. Waste with <260 ppm mercury--No specific treatment method is specified for hazardous wastes containing <260 ppm. However, RCRA regulations require that such wastes (other than RMERC residues) that exceed a TCLP mercury concentration of 0.20 mg/L be treated by a suitable method to meet the TCLP limit for mercury of 0.025 mg/L. RMERC residues must meet the TCLP value of {ge}0.20 mg/L, or be stabilized and meet the {ge}0.025 mg/L limit. Waste with {ge}260 ppm mercury--For hazardous wastes with mercury contaminant concentrations {ge}260 ppm and RCRA

  3. An Assessment of Using Vibrational Compaction of Calcined HLW and LLW in DWPF Canisters

    International Nuclear Information System (INIS)

    Yi, Yun-Bo; Amme, Robert C.; Shayer, Zeev

    2008-01-01

    Since 1963, the INEL has calcined almost 8 million gallons of liquid mixed waste and liquid high-level waste, converting it to some 1.1 million gallons of dry calcine (about 4275.0 m3), which consists of alumina-and zirconia-based calcine and zirconia-sodium blend calcine. In addition, if all existing and projected future liquid wastes are solidified, approximately 2,000 m3 of additional calcine will be produced primarily from sodium-bearing waste. Calcine is a more desirable material to store than liquid radioactive waste because it reduces volume, is much less corrosive, less chemically reactive, less mobile under most conditions, easier to monitor and more protective of human health and the environment. This paper describes the technical issue involved in the development of a feasible solution for further volume reduction of calcined nuclear waste for transportation and long term storage, using a standard DWPF canister. This will be accomplished by developing a process wherein the canisters are transported into a vibrational machine, for further volume reduction by about 35%. The random compaction experiments show that this volume reduction is achievable. The main goal of this paper is to demonstrate through computer modeling that it is feasible to use volume reduction vibrational machine without developing stress/strain forces that will weaken the canister integrity. Specifically, the paper presents preliminary results of the stress/strain analysis of the DWPF canister as a function of granular calcined height during the compaction and verifying that the integrity of the canister is not compromised. This preliminary study will lead to the development of better technology for safe compactions of nuclear waste that will have significant economical impact on nuclear waste storage and treatment. The preliminary results will guide us to find better solutions to the following questions: 1) What are the optimum locations and directions (vertical versus horizontal or

  4. The problems of hygienic classification of radioactive waste under restoration of contaminated areas

    International Nuclear Information System (INIS)

    Savkin, M.; Shandala, N.; Novikova, N.; Petukhova, E.; Shishkin, V.; Egorov, B.; Ziborov, A.

    2002-01-01

    Experience on restoration of contaminated areas in the past ten years reveals a specific problem in the general problem of solid radioactive waste management as a result of decontamination of the settlements. That specific problem concerns conventionally radioactive waste (CRW), which might be to some extent dangerous for human being. In the documents of IAEA and ICRP the approaches aimed at exemption or exclusion insignificant amount of radioactive wastes from regulatory control are actively being developed. In turn, Russia does not have so far either methodic or regulatory documents on management of very low level radioactive waste. Two approaches are considered in the paper under development of derived levels for CRW in case of restoration of contaminated areas. The first one is based on restriction of individual risk at level about 10 -6 per year (negligible level). The second one accounts for global man-made background and uses acceptable factor of excess of that background as a criterion.Under the first approach (restriction of individual risk) the lowest boundary of CRW is estimated to be equal to 3 Bq kg -1 for 239 Pu; 30 Bq kg -1 for 90 Sr; and 300 Bq kg -1 for 137 Cs, respectively. Those levels of specific activity approximately correspond to the areas contaminated by the above mentioned radionuclides 0.3 kBq m -2 , 3 kBq m -2 , and 30 kBq m -2 , respectively. Under the second approach if one accepts factor of 3 of excess of global man-made background, than the levels of specific activity will be 0.05 kBq m -2 for 239 Pu; 2.5 kBq m -2 for 90 Sr, and 7.2 kBq m -2 for 137 Cs. Comparison of the levels obtained according to the second approach shows that they will be several times lower than that according to the first approach. (author)

  5. Research and development program for transuranic-contaminated waste within the U.S. Energy Research and Development Administration

    International Nuclear Information System (INIS)

    Wolfe, R.A.

    1976-01-01

    This overview examines the research and development program that has been established within the U.S. Energy Research and Development Administration (ERDA) to develop the technology to treat transuranic-contaminated waste. Also considered is the waste expected within the total nuclear fuel cycle

  6. The risk implications of approaches to setting soil remediation goals at hazardous waste contaminated sites

    Energy Technology Data Exchange (ETDEWEB)

    Labieniec, Paula Ann [Carnegie Mellon Univ., Pittsburgh, PA (United States)

    1994-08-01

    An integrated exposure and carcinogenic risk assessment model for organic contamination in soil, SoilRisk, was developed and used for evaluating the risk implications of both site-specific and uniform-concentration approaches to setting soil remediation goals at hazardous-waste-contaminated sites. SoilRisk was applied to evaluate the uncertainty in the risk estimate due to uncertainty in site conditions at a representative site. It was also used to evaluate the variability in risk across a region of sites that can occur due to differences in site characteristics that affect contaminant transport and fate when a uniform concentration approach is used. In evaluating regional variability, Ross County, Ohio and the State of Ohio were used as examples. All analyses performed considered four contaminants (benzene, trichloroethylene (TCE), chlordane, and benzo[a]pyrene (BAP)) and four exposure scenarios (commercial, recreational and on- and offsite residential). Regardless of whether uncertainty in risk at a single site or variability in risk across sites was evaluated, the exposure scenario specified and the properties of the target contaminant had more influence than variance in site parameters on the resulting variance and magnitude of the risk estimate. In general, variance in risk was found to be greater for the relatively less degradable and more mobile of the chemicals studied (TCE and chlordane) than for benzene which is highly degradable and BAP which is very immobile in the subsurface.

  7. The risk implications of approaches to setting soil remediation goals at hazardous waste contaminated sites

    International Nuclear Information System (INIS)

    Labieniec, P.A.

    1994-08-01

    An integrated exposure and carcinogenic risk assessment model for organic contamination in soil, SoilRisk, was developed and used for evaluating the risk implications of both site-specific and uniform-concentration approaches to setting soil remediation goals at hazardous-waste-contaminated sites. SoilRisk was applied to evaluate the uncertainty in the risk estimate due to uncertainty in site conditions at a representative site. It was also used to evaluate the variability in risk across a region of sites that can occur due to differences in site characteristics that affect contaminant transport and fate when a uniform concentration approach is used. In evaluating regional variability, Ross County, Ohio and the State of Ohio were used as examples. All analyses performed considered four contaminants (benzene, trichloroethylene (TCE), chlordane, and benzo[a]pyrene (BAP)) and four exposure scenarios (commercial, recreational and on- and offsite residential). Regardless of whether uncertainty in risk at a single site or variability in risk across sites was evaluated, the exposure scenario specified and the properties of the target contaminant had more influence than variance in site parameters on the resulting variance and magnitude of the risk estimate. In general, variance in risk was found to be greater for the relatively less degradable and more mobile of the chemicals studied (TCE and chlordane) than for benzene which is highly degradable and BAP which is very immobile in the subsurface

  8. Effect of Canister Movement on Water Turbidity

    International Nuclear Information System (INIS)

    TRIMBLE, D.J.

    2000-01-01

    Requirements for evaluating the adherence characteristics of sludge on the fuel stored in the K East Basin and the effect of canister movement on basin water turbidity are documented in Briggs (1996). The results of the sludge adherence testing have been documented (Bergmann 1996). This report documents the results of the canister movement tests. The purpose of the canister movement tests was to characterize water turbidity under controlled canister movements (Briggs 1996). The tests were designed to evaluate methods for minimizing the plumes and controlling water turbidity during fuel movements leading to multi-canister overpack (MCO) loading. It was expected that the test data would provide qualitative visual information for use in the design of the fuel retrieval and water treatment systems. Video recordings of the tests were to be the only information collected

  9. Characterization and remediation of a mixed waste-contaminated site at Kirtland Air Force Base, New Mexico

    International Nuclear Information System (INIS)

    Johnston, J.W.; Thacker, M.S.; DeWitt, C.B.

    1997-01-01

    In the area of environmental restoration, one of the most challenging problems is the task of remediating mixed waste-contaminated sites. This paper discusses a successful Interim Corrective Measure (ICM) performed at a mixed waste-contaminated site on Kirtland Air Force Base (AFB) in Albuquerque, New Mexico. The site, known as RW-68, Cratering Area and Radium Dump/Slag Piles, was used during the late 1940s and early 1950s for the destruction and incineration of captured World War II aircraft. It contained 19 slag piles totaling approximately 150 tons of slag, ash, refractory brick, and metal debris. The piles were contaminated with radium-226 and RCRA-characteristic levels of heavy metals. Therefore, the piles were considered mixed waste. To eliminate the threat to human health and the environment, an ICM of removal, segregation, stabilization, and disposal was conducted from October through December 1996. Approximately 120 cubic yards (cu yds) of mixed waste, 188 cu yds of low-level radioactive-contaminated soil, 1 cu yd of low-level radioactive-contaminated debris, 5 cu yds of RCRA-characteristic hazardous waste, and 45 tons of nonhazardous debris were stabilized and disposed of during the ICM. To render the RCRA metals and radionuclides insoluble, stabilization was performed on the mixed and RCRA-characteristic waste streams. All stabilized material was subjected to TCLP analysis to verify it no longer exhibited RCRA-characteristic properties. Radiological and geophysical surveys were conducted concurrently with site remediation activities. These surveys provided real-time documentation of site conditions during each phase of the ICM and confirmed successful cleanup of the site. The three radioactive waste streams, stabilized mixed waste, low-level radioactive-contaminated soil, and low-level radioactive-contaminated debris, were disposed of at the Envirocare low-level radioactive disposal facility

  10. Heavy Metal Contamination of Soils around a Hospital Waste Incinerator Bottom Ash Dumps Site

    Directory of Open Access Journals (Sweden)

    M. Adama

    2016-01-01

    Full Text Available Waste incineration is the main waste management strategy used in treating hospital waste in many developing countries. However, the release of dioxins, POPs, and heavy metals in fly and bottom ash poses environmental and public health concerns. To determine heavy metal (Hg, Pb, Cd, Cr, and Ag in levels in incinerator bottom ash and soils 100 m around the incinerator bottom ash dump site, ash samples and surrounding soil samples were collected at 20 m, 40 m, 60 m, 80 m, 100 m, and 1,200 m from incinerator. These were analyzed using the absorption spectrophotometer method. The geoaccumulation (Igeo and pollution load indices (PLI were used to assess the level of heavy metal contamination of surrounding soils. The study revealed high concentrations in mg/kg for, Zn (16417.69, Pb (143.80, Cr (99.30, and Cd (7.54 in bottom ash and these were above allowable limits for disposal in landfill. The study also found soils within 60 m radius of the incinerator to be polluted with the metals. It is recommended that health care waste managers be educated on the implication of improper management of incinerator bottom ash and regulators monitor hospital waste incinerator sites.

  11. Applicability of molten salt oxidation to the destruction of actinide-contaminated wastes

    International Nuclear Information System (INIS)

    West, M.H.; Garcia, E.; Griego, W.J.; Court, D.B.; Rodriguez, L.

    1992-01-01

    A 1989 ban on incineration in the state of New Mexico caused cessation of actinide-contaminated cheesecloth, paper, and wood incineration within the Plutonium Facility (TA-55) at Los Alamos National Laboratory. Subsequently, plastic wipes were substituted for cheesecloth in the cleaning of glovebox interiors. However, waste minimization is not achieved by these measures since the wipes are discarded as Waste Isolation Pilot Plant certifiable wastes. After the ban was instituted, thermal decomposition of cheesecloth under argon at elevated temperature was examined and found satisfactory although scale of operation and speed were inferior to incineration. In 1991, the ban on incineration was lifted in New Mexico but Alamos has not chosen to pursue renewal of incineration at the Plutonium Facility. This paper reports that Los Alamos is looking from alternatives to incineration and thermal decomposition which are compatible with molten salt processing technology, historically a strength in actinide research at the Laboratory. Also, the technology must significantly reduce the volume of the waste upon treatment, i.e. waste minimization. Molten salt oxidation (MSO) has the promise of such a technology

  12. Heavy Metal Contamination of Soils around a Hospital Waste Incinerator Bottom Ash Dumps Site.

    Science.gov (United States)

    Adama, M; Esena, R; Fosu-Mensah, B; Yirenya-Tawiah, D

    2016-01-01

    Waste incineration is the main waste management strategy used in treating hospital waste in many developing countries. However, the release of dioxins, POPs, and heavy metals in fly and bottom ash poses environmental and public health concerns. To determine heavy metal (Hg, Pb, Cd, Cr, and Ag) in levels in incinerator bottom ash and soils 100 m around the incinerator bottom ash dump site, ash samples and surrounding soil samples were collected at 20 m, 40 m, 60 m, 80 m, 100 m, and 1,200 m from incinerator. These were analyzed using the absorption spectrophotometer method. The geoaccumulation (I geo) and pollution load indices (PLI) were used to assess the level of heavy metal contamination of surrounding soils. The study revealed high concentrations in mg/kg for, Zn (16417.69), Pb (143.80), Cr (99.30), and Cd (7.54) in bottom ash and these were above allowable limits for disposal in landfill. The study also found soils within 60 m radius of the incinerator to be polluted with the metals. It is recommended that health care waste managers be educated on the implication of improper management of incinerator bottom ash and regulators monitor hospital waste incinerator sites.

  13. Heavy Metal Contamination of Soils around a Hospital Waste Incinerator Bottom Ash Dumps Site

    Science.gov (United States)

    Adama, M.; Esena, R.; Fosu-Mensah, B.; Yirenya-Tawiah, D.

    2016-01-01

    Waste incineration is the main waste management strategy used in treating hospital waste in many developing countries. However, the release of dioxins, POPs, and heavy metals in fly and bottom ash poses environmental and public health concerns. To determine heavy metal (Hg, Pb, Cd, Cr, and Ag) in levels in incinerator bottom ash and soils 100 m around the incinerator bottom ash dump site, ash samples and surrounding soil samples were collected at 20 m, 40 m, 60 m, 80 m, 100 m, and 1,200 m from incinerator. These were analyzed using the absorption spectrophotometer method. The geoaccumulation (I geo) and pollution load indices (PLI) were used to assess the level of heavy metal contamination of surrounding soils. The study revealed high concentrations in mg/kg for, Zn (16417.69), Pb (143.80), Cr (99.30), and Cd (7.54) in bottom ash and these were above allowable limits for disposal in landfill. The study also found soils within 60 m radius of the incinerator to be polluted with the metals. It is recommended that health care waste managers be educated on the implication of improper management of incinerator bottom ash and regulators monitor hospital waste incinerator sites. PMID:27034685

  14. Fifth international conference on radioactive waste management and environmental remediation -- ICEM '95: Proceedings. Volume 2: Management of low-level waste and remediation of contaminated sites and facilities

    International Nuclear Information System (INIS)

    Slate, S.; Baker, R.; Benda, G.

    1995-01-01

    The objective of this conference is the broad international exchange of information on technologies, operations, management approaches, economics, and public policies in the critical areas of radioactive waste management and environmental remediation. The ICEM '95 technical program includes four parallel program tracks: Low/intermediate-level waste management; High-level waste, spent fuel, nuclear material management; Environmental remediation and facility D and D; and Major institutional issues in environmental management. Volume 2 contains approximately 200 papers divided into the following topical sections: Characterization of low and intermediate level waste; Treatment of low and intermediate level waste; LLW disposal and near-surface contaminant migration; Characterization and remediation of contaminated sites; and Decontamination and decommissioning technologies and experience. Papers have been processed separately for inclusion on the data base

  15. Nuclear Solid Waste Processing Design at the Idaho Spent Fuels Facility

    International Nuclear Information System (INIS)

    Dippre, M. A.

    2003-01-01

    A spent nuclear fuels (SNF) repackaging and storage facility was designed for the Idaho National Engineering and Environmental Laboratory (INEEL), with nuclear solid waste processing capability. Nuclear solid waste included contaminated or potentially contaminated spent fuel containers, associated hardware, machinery parts, light bulbs, tools, PPE, rags, swabs, tarps, weld rod, and HEPA filters. Design of the nuclear solid waste processing facilities included consideration of contractual, regulatory, ALARA (as low as reasonably achievable) exposure, economic, logistical, and space availability requirements. The design also included non-attended transfer methods between the fuel packaging area (FPA) (hot cell) and the waste processing area. A monitoring system was designed for use within the FPA of the facility, to pre-screen the most potentially contaminated fuel canister waste materials, according to contact- or non-contact-handled capability. Fuel canister waste materials which are not able to be contact-handled after attempted decontamination will be processed remotely and packaged within the FPA. Noncontact- handled materials processing includes size-reduction, as required to fit into INEEL permitted containers which will provide sufficient additional shielding to allow contact handling within the waste areas of the facility. The current design, which satisfied all of the requirements, employs mostly simple equipment and requires minimal use of customized components. The waste processing operation also minimizes operator exposure and operator attendance for equipment maintenance. Recently, discussions with the INEEL indicate that large canister waste materials can possibly be shipped to the burial facility without size-reduction. New waste containers would have to be designed to meet the drop tests required for transportation packages. The SNF waste processing facilities could then be highly simplified, resulting in capital equipment cost savings, operational

  16. Final Report: Characterization of Canister Mockup Weld Residual Stresses

    International Nuclear Information System (INIS)

    Enos, David; Bryan, Charles R.

    2016-01-01

    Stress corrosion cracking (SCC) of interim storage containers has been indicated as a high priority data gap by the Department of Energy (DOE) (Hanson et al., 2012), the Electric Power Research Institute (EPRI, 2011), the Nuclear Waste Technical Review Board (NWTRB, 2010a), and the Nuclear Regulatory Commission (NRC, 2012a, 2012b). Uncertainties exist in terms of the environmental conditions that prevail on the surface of the storage containers, the stress state within the container walls associated both with weldments as well as within the base metal itself, and the electrochemical properties of the storage containers themselves. The goal of the work described in this document is to determine the stress states that exists at various locations within a typical storage canister by evaluating the properties of a full-diameter cylindrical mockup of an interim storage canister. This mockup has been produced using the same manufacturing procedures as the majority of the fielded spent nuclear fuel interim storage canisters. This document describes the design and procurement of the mockup and the characterization of the stress state associated with various portions of the container. It also describes the cutting of the mockup into sections for further analyses, and a discussion of the potential impact of the results from the stress characterization effort.

  17. Final Report: Characterization of Canister Mockup Weld Residual Stresses

    Energy Technology Data Exchange (ETDEWEB)

    Enos, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-12-01

    Stress corrosion cracking (SCC) of interim storage containers has been indicated as a high priority data gap by the Department of Energy (DOE) (Hanson et al., 2012), the Electric Power Research Institute (EPRI, 2011), the Nuclear Waste Technical Review Board (NWTRB, 2010a), and the Nuclear Regulatory Commission (NRC, 2012a, 2012b). Uncertainties exist in terms of the environmental conditions that prevail on the surface of the storage containers, the stress state within the container walls associated both with weldments as well as within the base metal itself, and the electrochemical properties of the storage containers themselves. The goal of the work described in this document is to determine the stress states that exists at various locations within a typical storage canister by evaluating the properties of a full-diameter cylindrical mockup of an interim storage canister. This mockup has been produced using the same manufacturing procedures as the majority of the fielded spent nuclear fuel interim storage canisters. This document describes the design and procurement of the mockup and the characterization of the stress state associated with various portions of the container. It also describes the cutting of the mockup into sections for further analyses, and a discussion of the potential impact of the results from the stress characterization effort.

  18. The main rules regarding the management of solid waste and liquid effluent contaminated during use at nuclear medicine departments; Les principales regles de gestion des dechets solides et des effluents liquides contamines dans les services de medecine nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Boudouin, E. [Autorite de Surete Nucleaire, Direction des rayonnements ionisants et de la sante, 75 - Paris (France)

    2011-02-15

    This article describes the key requirements applicable to the management of contaminated medical waste and effluent from hospitals and health care centres, and more especially from nuclear medicine departments that use radionuclides for the purposes of diagnosis (in vivo or in vitro) or in patient treatment. It also presents the key management regulations, making a distinction between contaminated solid waste and contaminated liquid waste from such nuclear medicine departments. (author)

  19. New strains of oil-degrading microorganisms for treating contaminated soils and wastes

    Science.gov (United States)

    Muratova, A. Yu; Panchenko, L. V.; Semina, D. V.; Golubev, S. N.; Turkovskaya, O. V.

    2018-01-01

    Two new strains Achromobacter marplatensis101n and Acinetobacter sp. S-33, capable of degrading 49 and 46% of oil within 7 days were isolated, identified, and characterized. The application of A. marplatensis 101n in combination with ammonium nitrate (100 mg·kg-1) for 30 days of cultivation resulted in the degradation of 49% of the initial total petroleum hydrocarbon content (274 g·kg-1) in the original highly acid (pH 4.9) oil-contaminated waste. Up to 30% of oil sludge added to a liquid mineral medium at a concentration of 15% was degraded after 10 days of cultivation of A. marplatensis 101n. Application of yellow alfalfa (Medicago falcata L.) plants with Acinetobacter sp. S-33 for bioremediation of oil-sludge-contaminated soil improved the quality of cleanup in comparison with the bacterium- or plant-only treatment. Inoculation of Acinetobacter sp. S-33 increased the growth of both roots and shoots by more than 40%, and positively influenced the soil microflora. We conclude that the new oil-degrading strains, Acinetobacter sp. S-33 and A. marplatensis 101n, can serve as the basis for new bioremediation agents for the treatment of oil contaminated soils and waste.

  20. Characteristics and evaluation on heavy metal contamination in Changchun municipal waste landfill after closure.

    Science.gov (United States)

    Zhou, Xu-dan; Zhao, Chun-li; Qu, Tong-bao; Wang, Ying; Guo, Tai-jun; Sun, Xiao-gang

    2015-07-01

    In the present study, comprehensive investigation on the spot and typical investigation method were used to assess Mn, Zn, Pb, Cd, Cr, Ni, As and Cu level, pH value, organic matter, total nitrogen and total phosphorus contents in soil of Changchun municipal waste landfill. The results showed that soil in the closure area of Changchun municipal waste landfill was alkaline in nature and the average value of organic matter, total nitrogen and total phosphorus contents were lower than that in normal black soil in Changchun City of Jilin Province. Single factor indices of As, Pb and Cr content was > 1, where P(As) was 1.131, P(Pb) 1.061 and P(Cr) 1.092 mildly contaminated. In different sample spots but the same landfill time, the comprehensive Nemerow contamination indexes of 7a (5 #) and 7a (2 #) were P(2 comprehensive) = 1.176 and P(5 comprehensive) = 1.229. The performance value of of heavy metal contamination in soil was similar and there was a low ecological risk.

  1. Dose and risk assessment of norm Contaminated waste released from trench disposal facility

    International Nuclear Information System (INIS)

    Abdel Geleel, M.; Ramadan, A.B.; Tawfik, A.A.

    2005-01-01

    Oil and gas extraction and processing operations accumulate naturally occurring radioactive material (NORM) at concentrations above normal in by-product waste streams. The petroleum industry adopted methods for managing of NORM that are more restrictive than past practices and are likely to provide greater isolation of the radioactivity. Trench was used as a disposal facility for NORM contaminated wastes at one site of the petroleum industry in Egypt. The aim of this work is to calculate the risk and dose assessment received from trench disposal facility directly and after closure (1000 year). RESRAD computer code was used. The results indicated that the total effective dose (TED) received after direct closure of trench disposal facility was 7.7E-4 mSv/y while after 1000 years, it will he 3.4E-4. The health cancer risk after direct closure was 3.3E-8 while after 1000 years post closure it was 6E-8. Results of this assessment will help examine policy issues concerning different options and regulation of NORM contaminated waste generated by petroleum industry

  2. Physical properties of encapsulate spent fuel in canisters; Comportamiento fisico de las capsulas de almacenamiento

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-07-01

    Spent fuel and high-level wastes will be permanently stored in a deep geological repository (AGP). Prior to this, they will be encapsulated in canisters. The present report is dedicated to the study of such canisters under the different physical demands that they may undergo, be those in operating or accident conditions. The physical demands of interest include mechanical demands, both static and dynamic, and thermal demands. Consideration is given to the complete file of the canister, from the time when it is empty and without lid to the final conditions expected in the repository. Thermal analyses of canisters containing spent fuel are often carried out in two dimensions, some times with hypotheses of axial symmetry and some times using a plane transverse section through the centre of the canister. The results obtained in both types of analyses are compared here to those of complete three-dimensional analyses. The latter generate more reliable information about the temperatures that may be experienced by the canister and its contents; they also allow calibrating the errors embodied in the two-dimensional calculations. (Author)

  3. Topical safety analysis report for the transportation of the NUHOMS reg-sign dry shielded canister

    International Nuclear Information System (INIS)

    1993-08-01

    This Topical Safety Analysis Report (SAR) describes the design and the generic transportation licensing basis for utilizing the NUTECH HORIZONTAL MODULAR STORAGE (NUHOMS reg-sign) system dry shielded canister (DSC) containing twenty-four pressurized water reactor (PWR) spent fuel assemblies (SFA) in conjunction with a conceptually designed Transportation Cask. This SAR documents the design qualification of the NUHOMS reg-sign DSC as an integral part of a 10CFR71 Fissile Material Class III, Type B(M) Transportation Package. The package consists of the canister and a conceptual transportation cask (NUHOMS reg-sign Transportation Cask) with impact limiters. Engineering analysis is performed for the canister to confirm that the existing canister design complies with 10CFR71 transportation requirements. Evaluations and/or analyses is performed for criticality safety, shielding, structural, and thermal performance. Detailed engineering analysis for the transportation cask will be submitted in a future SAR requesting 10CFR71 certification of the complete waste package. Transportation operational considerations describe various operational aspects of the canister/transportation cask system. operational sequences are developed for canister transfer from storage to the transportation cask and interfaces with the cask auxiliary equipment for on- and off-site transport

  4. Fifth in situ vitrification engineering-scale test of simulated INEL buried waste sites

    International Nuclear Information System (INIS)

    Bergsman, T.M.; Shade, J.W.; Farnsworth, R.K.

    1992-06-01

    In September 1990, an engineering-scale in situ vitrification (ISV) test was conducted on sealed canisters containing a combined mixture of buried waste materials expected to be present at the Idaho National Engineering Laboratory (INEL) Subsurface Disposal Area (SDA). The test was part of a Pacific Northwest Laboratory (PNL) program to assist INEL in treatability studies of the potential application of ISV to mixed transuranic wastes at the INEL SDA. The purpose of this test was to determine the effect of a close-packed layer of sealed containers on ISV processing performance. Specific objectives included determining (1) the effect of releases from sealed containers on hood plenum pressure and temperature, (2) the release pressure ad temperatures of the sealed canisters, (3) the relationships between canister depressurization and melt encapsulation, (4) the resulting glass and soil quality, (5) the potential effects of thermal transport due to a canister layer, (6) the effects on particle entrainment of differing angles of approach for the ISV melt front, and (7) the effects of these canisters on the volatilization of voltatile and semivolatile contaminants into the hood plenum

  5. Corrosion test plan to guide canister material selection and design for a tuff repository

    International Nuclear Information System (INIS)

    McCright, R.D.; van Konynenburg, R.A.; Ballou, L.B.

    1983-11-01

    Corrosion rates and the mode of corrosion attack form a most important basis for selection of canister materials and design of a nuclear waste package. Type 304L stainless steel was selected as the reference material for canister fabrication because of its generally excellent corrosion resistance in water, steam and air. However, 304L may be susceptible to localized and stress-assisted forms of corrosion under certain conditions. Alternative alloys are also investigated; these alloys were chosen because of their improved resistance to these forms of corrosion. The fabrication and welding processes, as well as the glass pouring operation for defense and commercial high-level wastes, may influence the susceptibility of the canister to localized and stress forms of corrosion. 12 references, 2 figures, 4 tables

  6. CANISTER HANDLING FACILITY DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    J.F. Beesley

    2005-04-21

    The purpose of this facility description document (FDD) is to establish requirements and associated bases that drive the design of the Canister Handling Facility (CHF), which will allow the design effort to proceed to license application. This FDD will be revised at strategic points as the design matures. This FDD identifies the requirements and describes the facility design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This FDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This FDD is part of an iterative design process. It leads the design process with regard to the flowdown of upper tier requirements onto the facility. Knowledge of these requirements is essential in performing the design process. The FDD follows the design with regard to the description of the facility. The description provided in this FDD reflects the current results of the design process.

  7. CANISTER HANDLING FACILITY DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    Beesley. J.F.

    2005-01-01

    The purpose of this facility description document (FDD) is to establish requirements and associated bases that drive the design of the Canister Handling Facility (CHF), which will allow the design effort to proceed to license application. This FDD will be revised at strategic points as the design matures. This FDD identifies the requirements and describes the facility design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This FDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This FDD is part of an iterative design process. It leads the design process with regard to the flowdown of upper tier requirements onto the facility. Knowledge of these requirements is essential in performing the design process. The FDD follows the design with regard to the description of the facility. The description provided in this FDD reflects the current results of the design process

  8. Settlement of Canisters with smectite clay envelopes in deposition holes

    International Nuclear Information System (INIS)

    Pusch, R.

    1986-12-01

    Settlement of canisters containing radioactive waste and being surrounded by dense smectite clay is caused by the stresses and heat induced in the clay. Consolidation by water expulsion of the clay underlying a model canister with 5 cm diameter and 30 cm length would theoretically account for a maximum finite settlement of about 70 my m in a few weeks, while shear-induced creep would yield a settlement of only a few microns in the same time period. These predictions were checked by running a laboratory test in which a dead load of 80 kg was applied to a small cylindrical copper canister embedded in Na bentonite. The settlement, which increased in proportion to log time, turned out to be about 6 my m in the first 2.5 months. After the first loading period at room temperature, heating to 50 degrees C and, after a 4 months long 'room temperature' period, to 70 degrees C took place. This cycling gave strong, instant settlement and upheaval because of the different thermal expansion of the interacting components of the system. After the development of constant temperature conditions in the entire system and completion of the consolidation or expansion that followed from the thermo-mechanical interactions, the settlement proceeded at a rather high rate at 70 degrees C, still following a log time creep law, but with somewhat stronger retardation. At room temperature, i.e. in the post-heating periods, the settlement seemed to cease, on the other hand. The conclusion from the study is that the canister movements under isothermal conditions were in accordance with the log t-type creep settlement that was predicted in theoretical grounds. Pre-heating and low stresses may account for extraordinary retardation of the settlement. (author)

  9. Analysis of probability of defects in the disposal canisters

    International Nuclear Information System (INIS)

    Holmberg, J.-E.; Kuusela, P.

    2011-06-01

    This report presents a probability model for the reliability of the spent nuclear waste final disposal canister. Reliability means here that the welding of the canister lid has no critical defects from the long-term safety point of view. From the reliability point of view, both the reliability of the welding process (that no critical defects will be born) and the non-destructive testing (NDT) process (all critical defects will be detected) are equally important. In the probability model, critical defects in a weld were simplified into a few types. Also the possibility of human errors in the NDT process was taken into account in a simple manner. At this moment there is very little representative data to determine the reliability of welding and also the data on NDT is not well suited for the needs of this study. Therefore calculations presented here are based on expert judgements and on several assumptions that have not been verified yet. The Bayesian probability model shows the importance of the uncertainty in the estimation of the reliability parameters. The effect of uncertainty is that the probability distribution of the number of defective canisters becomes flat for larger numbers of canisters compared to the binomial probability distribution in case of known parameter values. In order to reduce the uncertainty, more information is needed from both the reliability of the welding and NDT processes. It would also be important to analyse the role of human factors in these processes since their role is not reflected in typical test data which is used to estimate 'normal process variation'.The reported model should be seen as a tool to quantify the roles of different methods and procedures in the weld inspection process. (orig.)

  10. Application of the TEMPEST computer code to canister-filling heat transfer problems

    International Nuclear Information System (INIS)

    Farnsworth, R.K.; Faletti, D.W.; Budden, M.J.

    1988-03-01

    Pacific Northwest Laboratory (PNL) researchers used the TEMPEST computer code to simulate thermal cooldown behavior of nuclear waste glass after it was poured into steel canisters for long-term storage. The objective of this work was to determine the accuracy and applicability of the TEMPEST code when used to compute canister thermal histories. First, experimental data were obtained to provide the basis for comparing TEMPEST-generated predictions. Five canisters were instrumented with appropriately located radial and axial thermocouples. The canister were filled using the pilot-scale ceramic melter (PSCM) at PNL. Each canister was filled in either a continous or a batch filling mode. One of the canisters was also filled within a turntable simulant (a group of cylindrical shells with heat transfer resistances similar to those in an actual melter turntable). This was necessary to provide a basis for assessing the ability of the TEMPEST code to also model the transient cooling of canisters in a melter turntable. The continous-fill model, Version M, was found to predict temperatures with more accuracy. The turntable simulant experiment demonstrated that TEMPEST can adequately model the asymmetric temperature field caused by the turntable geometry. Further, TEMPEST can acceptably predict the canister cooling history within a turntable, despite code limitations in computing simultaneous radiation and convection heat transfer between shells, along with uncertainty in stainless-steel surface emissivities. Based on the successful performance of TEMPEST Version M, development was initiated to incorporate 1) full viscous glass convection, 2) a dynamically adaptive grid that automatically follows the glass/air interface throughout the transient, and 3) a full enclosure radiation model to allow radiation heat transfer to non-nearest neighbor cells. 5 refs., 47 figs., 17 tabs

  11. Potency of Centrocema pubescence, Calopogonium mucunoides, and Micania cordata for cleaning metal contaminants of gold mines waste

    Directory of Open Access Journals (Sweden)

    NURIL HIDAYATI

    2006-01-01

    Full Text Available Based on some findings that some plants are tolerant to contaminated media, this research was conducted to study more thoroughly about characters and potencies of some of them as hyperaccumulators. Three of the most tolerant plants were studied in this research i.e Centrocema pubescence, Calopogonium mucunoides,and Micania cordata. The plants were grown in different waste media, i.e. tailing from PT. Aneka Tambang (ANTAM and people mine waste. Both waste have different characters, physically and chemically. Waste of PT ANTAM major contaminant was cianide (Cn whereas people mine waste major contaminant was mercury (Hg. This different characters resulted in different plant responses. The plants grown under PT ANTAM waste media gave better performance than that grown under people mine waste media. The most tolerant species was C. pubescence followed by M. cordata and C. mucunoides. Ability in metal accumulation of C. mucunoides was the highest, followed by M. cordata and C. pubescence.The results raised some-prospects for phytoremediation technology for rehabilitating contaminated mined lands.

  12. Establishing indices for groundwater contamination risk assessment in the vicinity of hazardous waste landfills in China.

    Science.gov (United States)

    Li, Ying; Li, Jinhui; Chen, Shusheng; Diao, Weihua

    2012-06-01

    Groundwater contamination by leachate is the most damaging environmental impact over the entire life of a hazardous waste landfill (HWL). With the number of HWL facilities in China rapidly increasing, and considering the poor status of environmental risk management, it is imperative that effective environmental risk management methods be implemented. A risk assessment indices system for HWL groundwater contamination is here proposed, which can simplify the risk assessment procedure and make it more user-friendly. The assessment framework and indices were drawn from five aspects: source term, underground media, leachate properties, risk receptors and landfill management quality, and a risk assessment indices system consisting of 38 cardinal indicators was established. Comparison with multimedia models revealed that the proposed indices system was integrated and quantitative, that input data for it could be easily collected, and that it could be widely used for environmental risk assessment (ERA) in China. Crown Copyright © 2012. Published by Elsevier Ltd. All rights reserved.

  13. Contaminant transport modelling in tidal influenced water body for low level liquid waste discharge out

    International Nuclear Information System (INIS)

    Singh, Sanjay; Naidu, Velamala Simhadri

    2018-01-01

    Low level liquid waste is generated from nuclear reactor operation and reprocessing of spent fuel. This waste is discharged into the water body after removing bulk of its radioactivity. Dispersion of contaminant mainly depends on location of outfall and hydrodynamics of water body. For radiological impact assessment, in most of the analytical formulations, source term is taken as continuous release. However, this may not be always true as the water level is influenced by tidal movement and the selected outfall may come under intertidal zone in due course of the tidal cycle. To understand these phenomena, a case study has been carried out to evaluate hydrodynamic characteristics and dilution potential of outfall located in inter-tidal zone using numerical modelling

  14. Thermal decomposition of woody wastes contaminated with radioactive materials using externally-heated horizontal kiln

    International Nuclear Information System (INIS)

    Iwasaki, Toshiyuki; Kato, Shigeru; Yamasaki, Akihiro; Ito, Takuya; Suzuki, Seiichi; Kojima, Toshinori; Kodera, Yoichi; Hatta, Akimichi; Kikuzato, Masahiro

    2015-01-01

    Thermal decomposition experiments of woody wastes contaminated with radioactive materials were conducted using an externally-heated horizontal kiln in the work area for segregation of disaster wastes at Hirono Town, Futaba County, Fukushima Prefecture. Radioactivity was not detected in gaseous products of thermal decomposition at 923 K and 1123 K after passage through a trap filled with activated carbon. The contents of radioactive cesium ( 134 Cs and 137 Cs) were measured in the solid and liquid products of the thermal decomposition experiments and in the residues in the kiln after all of the experiments. Although a trace amount of radioactive cesium was found in the washing trap during the start-up period of operation at 923 K, most of the cesium remained in the char, including the residues in the kiln. These results suggest that most of the radioactive cesium is trapped in char particles and is not emitted in gaseous form. (author)

  15. Green waste compost as an amendment during induced phytoextraction of mercury-contaminated soil.

    Science.gov (United States)

    Smolinska, Beata

    2015-03-01

    Phytoextraction of mercury-contaminated soils is a new strategy that consists of using the higher plants to make the soil contaminant nontoxic. The main problem that occurs during the process is the low solubility and bioavailability of mercury in soil. Therefore, some soil amendments can be used to increase the efficiency of the Hg phytoextraction process. The aim of the investigation was to use the commercial compost from municipal green wastes to increase the efficiency of phytoextraction of mercury-contaminated soil by Lepidium sativum L. plants and determine the leaching of Hg after compost amendment. The result of the study showed that Hg can be accumulated by L. sativum L. The application of compost increased both the accumulation by whole plant and translocation of Hg to shoots. Compost did not affect the plant biomass and its biometric parameters. Application of compost to the soil decreased the leaching of mercury in both acidic and neutral solutions regardless of growing medium composition and time of analysis. Due to Hg accumulation and translocation as well as its potential leaching in acidic and neutral solution, compost can be recommended as a soil amendment during the phytoextraction of mercury-contaminated soil.

  16. Gelatinous soil barrier for reducing contaminant emissions at waste-disposal sites

    International Nuclear Information System (INIS)

    Opitz, B.E.; Martin, W.J.; Sherwood, D.R.

    1982-09-01

    The milling of uranium ore produces large quantities of waste (mill tailings) that are being deposited in earthen pits or repositories. These wastes, which remain potentially hazardous for long time periods, may reach the biosphere at levels greater than those allowed by the Environmental Protection Agency (EPA). For example, the leachates associated with these wastes contain numerous radionuclides and toxic trace metals at levels 10 2 to 10 4 greater than allowable for drinking water based on EPA Primary Drinking Water Standards. As a result, technologies must be developed to ensure that such wastes will not reach the biosphere at hazardous levels. Under sponsorship of the Department of Energy's Uranium Mill Tailings Remedial Action Program (UMTRAP), Pacific Northwest Laboratory (PNL) has investigated the use of engineered barriers for use as liners and covers for waste containment. Results of these investigations have led to the development of a low permeable, multilayer earthen barrier that effectively reduces contaminant loss from waste disposal sites. The multilayer earth barrier was developed as an alternative to clay liner or cover schemes for use in areas where clays were not locally available and must be shipped to the disposal site. The barrier layer is comprised of 90% locally available materials whose liner or cover properties are enhanced by the addition of a gelatinous precipitate which entrains moisture into the cover's air-filled pore spaces, blocking the pathways through which gas would otherwise diffuse into the atmosphere or through which moisture would migrate into the ground. In field verification tests, the earthen seal reduced radon gas emissions by 95 to 99% over prior release rates with measured permeabilities on the order of 10 - 9 cm/s

  17. Application of autonomous robotics to surveillance of waste storage containers for radioactive surface contamination

    International Nuclear Information System (INIS)

    Sweeney, F.J.; Beckerman, M.; Butler, P.L.; Jones, J.P.; Reister, D.B.

    1991-01-01

    This paper describes a proof-of-principal demonstration performed with the HERMIES-III mobile robot to automate the inspection of waste storage drums for radioactive surface contamination and thereby reduce the human burden of operating a robot and worker exposure to potentially hazardous environments. Software and hardware for the demonstration were developed by a team consisting of Oak Ridge National Laboratory, and the Universities of Florida, Michigan, Tennessee, and Texas. Robot navigation, machine vision, manipulator control, parallel processing and human-machine interface techniques developed by the team were demonstrated utilizing advanced computer architectures. The demonstration consists of over 100,000 lines of computer code executing on nine computers

  18. Levels for the specific activity at disposing low-level contaminated municipal wastes

    International Nuclear Information System (INIS)

    Poschner, J.; Schaller, G.

    1993-12-01

    Using radioecological models, nuclide-specific ''basic values'' were established for the specific activity of radioactive contaminated waste. These basic values were determined in order to avoid dose equivalents of more than 10 μSv/a (de minimis concept) when disposing the waste in conventional waste deposits or burning it in incineration plants. The basic values vary within 16 orders of magnitude, ranging between 4.3E-02 Bq/g for Cm-250 and 2.2E+14 Bq/g for Po-212. About 82% of the basic values are between 1 Bq/g und 1000 Bq/g. The critical exposure for most of the radionuclides is that from direct radiation or inhalation at the slag-deposit. For 54% of the radionuclides the basic value for the specific activity of waste is lower at least by a factor of 10 than the 10 -4 -fold value per gram of the release limit (German Radiation Protection Ordinance). Appropriate nuclide-specific and dose-related limits were derived from these basic values. The respective limits are ranging between 1E-01 Bq/g and 1E+07 Bq/g. About 95% of the limits are between 1 Bq/g and 1000 Bq/g. (orig./HP) [de

  19. In Vivo Cytogenotoxicity and Oxidative Stress Induced by Electronic Waste Leachate and Contaminated Well Water

    Directory of Open Access Journals (Sweden)

    Adeyinka M. Gbadebo

    2013-07-01

    Full Text Available Environmental, plant and animal exposure to hazardous substances from electronic wastes (e-wastes in Nigeria is increasing. In this study, the potential cytogenotoxicity of e-wastes leachate and contaminated well water samples obtained from Alaba International Electronic Market in Lagos, Nigeria, using induction of chromosome and root growth anomalies in Allium cepa, and micronucleus (MN in peripheral erythrocytes of Clarias gariepinus, was evaluated. The possible cause of DNA damage via the assessments of liver malondialdehyde (MDA, catalase (CAT, reduced glutathione (GSH and superoxide dismutase (SOD as indicators of oxidative stress in mice was also investigated. There was significant (p < 0.05 inhibition of root growth and mitosis in A. cepa. Cytological aberrations such as spindle disturbance, C-mitosis and binucleated cells, and morphological alterations like tumor and twisting roots were also induced. There was concentration-dependent, significant (p < 0.05 induction of micronucleated erythrocytes and nuclear abnormalities such as blebbed nuclei and binucleated erythrocytes in C. gariepinus. A significant increase (p < 0.001 in CAT, GSH and MDA with concomitant decrease in SOD concentrations were observed in the treated mice. Pb, As, Cu, Cr, and Cd analyzed in the tested samples contributed significantly to these observations. This shows that the well water samples and leachate contained substances capable of inducing somatic mutation and oxidative stress in living cells; and this is of health importance in countries with risk of e-wastes exposure.

  20. Risk assessment of particle dispersion and trace element contamination from mine-waste dumps.

    Science.gov (United States)

    Romero, Antonio; González, Isabel; Martín, José María; Vázquez, María Auxiliadora; Ortiz, Pilar

    2015-04-01

    In this study, a model to delimit risk zones influenced by atmospheric particle dispersion from mine-waste dumps is developed to assess their influence on the soil and the population according to the concentration of trace elements in the waste. The model is applied to the Riotinto Mine (in SW Spain), which has a long history of mining and heavy land contamination. The waste materials are separated into three clusters according to the mapping, mineralogy, and geochemical classification using cluster analysis. Two of the clusters are composed of slag, fresh pyrite, and roasted pyrite ashes, which may contain high concentrations of trace elements (e.g., >1 % As or >4 % Pb). The average pollution load index (PLI) calculated for As, Cd, Co, Cu, Pb, Tl, and Zn versus the baseline of the regional soil is 19. The other cluster is primarily composed of sterile rocks and ochreous tailings, and the average PLI is 3. The combination of particle dispersion calculated by a Gaussian model, the PLI, the surface area of each waste and the wind direction is used to develop a risk-assessment model with Geographic Information System GIS software. The zone of high risk can affect the agricultural soil and the population in the study area, particularly if mining activity is restarted in the near future. This model can be applied to spatial planning and environmental protection if the information is complemented with atmospheric particulate matter studies.

  1. Hanford Tank 241-C-106: Impact of Cement Reactions on Release of Contaminants from Residual Waste

    International Nuclear Information System (INIS)

    Deutsch, William J.; Krupka, Kenneth M.; Lindberg, Michael J.; Cantrell, Kirk J.; Brown, Christopher F.; Schaef, Herbert T.

    2006-01-01

    The CH2M HILL Hanford Group, Inc. (CH2M HILL) is producing risk/performance assessments to support the closure of single-shell tanks at the U.S. Department of Energy's Hanford Site. As part of this effort, staff at Pacific Northwest National Laboratory were asked to develop release models for contaminants of concern that are present in residual sludge remaining in tank 241-C-106 (C-106) after final retrieval of waste from the tank. Initial work to produce release models was conducted on residual tank sludge using pure water as the leaching agent. The results were reported in an earlier report. The decision has now been made to close the tanks after waste retrieval with a cementitious grout to minimize infiltration and maintain the physical integrity of the tanks. This report describes testing of the residual waste with a leaching solution that simulates the composition of water passing through the grout and contacting the residual waste at the bottom of the tank.

  2. Co-combustion of gasified contaminated waste wood in a coal fired power plant

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    This project demonstrates the technical and economical feasibility of the producing and cofiring of product gas from demolition waste wood. For this purpose LCV product gas is generated in an atmospheric circulating fluidized bed (CFB) gasification plant, cooled and cleaned and transported to the boiler of a 600 MWe pulverized coal fired power plant. Gas cooling and cleaning takes place in a waste heat boiler and a multi stage wet gas cleaning train. Steam raised in the waste heat boiler is exported to the power plant. On an annual basis 70,000 tons of steam coal are substituted by 150,000 tons of contaminated demolition waste wood (50,000 tons oil equivalent), resulting in a net CO2 emission reduction of 170,000 tons per year, while concurrently generating 205 GWh of electrical power. The wood gasification plant was built by NV EPZ (now incorporated in Essent Energi BV) for Amergas BV, now a 100% subsidiary of Essent Energie BV. The gasification plant is located at the Amer Power Station of NV EPZ Production (now Essent Generation) at Geertruidenberg, The Netherlands. Demonstrating several important design features in wood gasification, the plant started hot service in the Spring of 2000, with first gasification accomplished in the Summer of 2000 and is currently being optimized. (au)

  3. Latest movements associated with radioactive contamination and disaster waste management (2)

    International Nuclear Information System (INIS)

    Omura, Tomomi

    2012-01-01

    As for the radioactive contamination countermeasures and disaster waste countermeasures taken for the accidents of the Great East Japan Earthquake and the Fukushima Daiichi Nuclear Station of Tokyo Electric Power Company, this paper introduces in the digest version the following movements from mid-April to May 15, 2012. (1) Radioactive substance countermeasures such as decontamination. (a) Decontamination operations under direct control of the Ministry of the Environment, (b) Establishment of compensation benchmarks by the Ministry of the Environment for the garden plants and land use in Special Decontamination Area, (c) Publication of technical guidelines by the Ministry of Agriculture, Forestry and Fisheries, on the removal and diffusion suppression of radioactive substances in forests, (d) Announcement of research center development / promotion idea by the government in the policy making for Fukushima reconstruction, (e) Request of the government for the interim storage facility site in the opinion exchange meetings in Futaba district towns and villages in Fukushima Prefecture, (f) Announcement of radioactive substance forecast map in Fukushima City for the first time by the government, and (g) Action plan development at the Health Anxiety Countermeasure Coordination Council for nuclear victims. (2) Disaster waste countermeasures. (a) Introduction of challenges in each of Miyagi Prefecture and Iwate Prefecture on the acceleration of the secondary temporary storage field development for disaster waste treatment, and (b) Introduction of progress in new interim incinerator construction plan for disaster waste treatment in Fukushima Prefecture. (O.A.)

  4. Water hyacinth for phytoremediation of radioactive waste simulate contaminated with cesium and cobalt radionuclides

    International Nuclear Information System (INIS)

    Saleh, H.M.

    2012-01-01

    Highlights: ► Phytoremediation of radioactive