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Sample records for calcined high-level waste

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    T. R. Thomas; A. K. Herbst

    2005-06-01

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

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

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1993-06-01

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

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

    International Nuclear Information System (INIS)

    Freeby, W.A.

    1975-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-05-01

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

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

    International Nuclear Information System (INIS)

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

    1997-05-01

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

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  9. Development of Concentration and Calcination Technology For High Level Liquid Waste

    International Nuclear Information System (INIS)

    Pande, D.P.

    2006-01-01

    The concentrated medium and high-level liquid radio chemicals effluents contain nitric acid, water along with the dissolved chemicals including the nitrates of the radio nuclides. High level liquid waste contain mainly nitrates of cesium, strontium, cerium, zirconium, chromium, barium, calcium, cobalt, copper, pickle, iron etc. and other fission products. This concentrated solution requires further evaporation, dehydration, drying and decomposition in temperature range of 150 to 700 deg. C. The addition of the calcined solids in vitrification pot, instead of liquid feed, helps to avoid low temperature zone because the vaporization of the liquid and decomposition of nitrates do not take place inside the melter. In our work Differential and thermo gravimetric studies has been carried out in the various stages of thermal treatment including drying, dehydration and conversion to oxide forms. Experimental studies were done to characterize the chemicals present in high-level radioactive waste. A Rotary Ball Kiln Calciner was used for development of the process because this is amenable for continuous operation and moderately good heat transfer can be achieved inside the kiln. This also has minimum secondary waste and off gases generation. The Rotary Ball Kiln Calciner Demonstration facility system was designed and installed for the demonstration of calcination process. The Rotary Ball Kiln Calciner is a slowly rotating slightly inclined horizontal tube that is externally heated by means of electric resistance heating. The liquid feed is sprayed onto the moving bed of metal balls in a slowly rotating calciner by a peristaltic type-metering pump. The vaporization of the liquid occurs in the pre-calcination zone due to counter current flow of hot gases. The dehydration and denitration of the solids occurs in the calcination zone, which is externally heated by electrical furnace. The calcined powder is cooled in the post calcination portion. It has been demonstrated that the

  10. Pyrochemical separation of radioactive components from inert materials in ICPP high-level calcined waste

    International Nuclear Information System (INIS)

    Del Debbio, J.A.; Nelson, L.O.; Todd, T.A.

    1995-05-01

    Since 1963, calcination of aqueous wastes from reprocessing of DOE-owned spent nuclear fuels has resulted in the accumulation of approximately 3800 m 3 of high-level waste (HLW) at the Idaho Chemical Processing Plant (ICPP). The waste is in the form of a granular solid called calcine and is stored on site in stainless steel bins which are encased in concrete. Due to the leachability of 137 Cs and 90 Sr and possibly other radioactive components, the calcine is not suitable for final disposal. Hence, a process to immobilize calcine in glass is being developed. Since radioactive components represent less than 1 wt % of the calcine, separation of actinides and fission products from inert components is being considered to reduce the volume of HLW requiring final disposal. Current estimates indicate that compared to direct vitrification, a volume reduction factor of 10 could result in significant cost savings. Aqueous processes, which involve calcine dissolution in nitric acid followed by separation of actinide and fission products by solvent extraction and ion exchange methods, are being developed. Pyrochemical separation methods, which generate small volumes of aqueous wastes and do not require calcine dissolution, have been evaluated as alternatives to aqueous processes. This report describes three proposed pyrochemical flowsheets and presents the results of experimental studies conducted to evaluate their feasibility. The information presented is a consolidation of three reports, which should be consulted for experimental details

  11. Spray Calciner/In-Can Melter high-level waste solidification technical manual

    International Nuclear Information System (INIS)

    Larson, D.E.

    1980-09-01

    This technical manual summarizes process and equipment technology developed at Pacific Northwest Laboratory over the last 20 years for vitrification of high-level liquid waste by the Spray Calciner/In-Can Melter process. Pacific Northwest Laboratory experience includes process development and demonstration in laboratory-, pilot-, and full-scale equipment using nonradioactive synthetic wastes. Also, laboratory- and pilot-scale process demonstrations have been conducted using actual high-level radioactive wastes. In the course of process development, more than 26 tonnes of borosilicate glass have been produced in 75 canisters. Four of these canisters contained radioactive waste glass. The associated process and glass chemistry is discussed. Technology areas described include calciner feed treatment and techniques, calcination, vitrification, off-gas treatment, glass containment (the canister), and waste glass chemistry. Areas of optimization and site-specific development that would be needed to adapt this base technology for specific plant application are indicated. A conceptual Spray Calciner/In-Can Melter system design and analyses are provided in the manual to assist prospective users in evaluating the process for plant application, to provide equipment design information, and to supply information for safety analyses and environmental reports. The base (generic) technology for the Spray Calciner/In-Can Melter process has been developed to a point at which it is ready for plant application

  12. Pyrochemical treatment of Idaho Chemical Processing Plant high-level waste calcine

    International Nuclear Information System (INIS)

    Todd, T.A.; DelDebbio, J.A.; Nelson, L.O.; Sharpsten, M.R.

    1993-01-01

    The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Engineering Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1951 to recover uranium, krypton-85, and isolated fission products for interim treatment and immobilization. The acidic radioactive high-level liquid waste (HLLW) is routinely stored in stainless steel tanks and then, since 1963, calcined to form a dry granular solid. The resulting high-level waste (HLW) calcine is stored in seismically hardened stainless steel bins that are housed in underground concrete vaults. A research and development program has been established to determine the feasibility of treating ICPP HLW calcine using pyrochemical technology.This technology is described

  13. Applied laboratory research of high-level waste denitration and calcination technologies

    International Nuclear Information System (INIS)

    Napravnik, J.

    1977-01-01

    Denitration and calcination processes are assessed for model solutions of high-level radioactive wastes. The kinetics was studied of the reaction of HNO 3 with HCOOH with respect to the final composition of the gaseous product. A survey is presented of used denitration agents and of reaction processes. Calcination was studied both as associated with denitration in a single technological step and separately. Also studied was the pyrolysis and chemical decomposition of sodium nitrate which forms an indecomposable melt in the temperature region of 320 to 850 degC under normal conditions. Based on the experiments a laboratory unit was designed and produced for the denitration and calcination of model solutions of high-level radioactive wastes operating in a temperature range of 100 to 550 degC with a capacity of 1000 ml/h. A boiler type stirred evaporator with electric heating (3 kW) was chosen for the denitration unit while a vertical calcinator modified from a film evaporator with a thermal input of 4 kW was chosen for the calcination unit. (B.S.)

  14. Design and performance of atomizing nozzles for spray calcination of high-level wastes

    International Nuclear Information System (INIS)

    Miller, F.A.; Stout, L.A.

    1981-05-01

    A key aspect of high-level liquid-waste spray calcination is waste-feed atomization by using air atomizing nozzles. Atomization substantially increases the heat transfer area of the waste solution, which enhances rapid drying. Experience from the spray-calciner operations has demonstrated that nozzle flow conditions that produce 70-μ median-volume-diameter or smaller spray droplets are required for small-scale spray calciners (drying capacity less than 80 L/h). For large-scale calciners (drying capacity greater than 300 L/h), nozzle flow conditions that produce 100-μ median-volume-diameter or smaller spray droplets are required. Mass flow ratios of 0.2 to 0.4, depending on nozzle size, are required for proper operation of internal-mix atomizing nozzles. Both internal-mix and external-mix nozzles have been tested at PNL. Due to the lower airflow requirements and fewer large droplets produced, the internal-mix nozzle has been chosen for primary development in the spray calciner program at PNL. Several nozzle air-cap materials for internal-mix nozzles have been tested for wear resistance. Results show that nozzle air caps of stainless steel and Cer-vit (a machineable glass ceramic) are suceptible to rapid wear by abrasive slurries, whereas air caps of alumina and reaction-bonded silicon nitride show only slow wear. Longer-term testing is necessary to determine more accurately the actual frequency of nozzle replacement. Atomizing nozzle air caps of alumina are subject to fracture from thermal shock, whereas air caps of silicon nitride and Cer-vit are not. Fractured nozzles are held in place by the air-cap retaining ring and continue to atomize satisfactorily. Therefore, fractures caused by thermal shocking do not necessarily result in nozzle failure

  15. Behavior of radioactive iodine and technetium in the spray calcination of high-level waste

    Science.gov (United States)

    Knox, C. A.; Farnsworth, R. K.

    1981-08-01

    The Remote Laboratory-Scale Waste Treatment Facility (RLSWTF) was designed and built as a part of the High-Level Waste Immobilization Program (now the High-Level Waste Process Development Program) at the Pacific Northwest Laboratory. In facility, installed in a radiochemical cell, is described in which installed in a radiochemical cell is described in which small volumes of radioactive liquid wastes can be solidified, the process off gas can be analyzed, and the methods for decontaminating this off gas can be tested. During the spray calcination of commercial high-level liquid waste spiked with Tc-99 and I-131 and 31 wt% loss of I-131 past the sintered-metal filters. These filters and venturi scrubber were very efficient in removing particulates and Tc-99 from the the off-gas stream. Liquid scrubbers were not efficient in removing I-131 as 25% of the total lost went to the building off-gas system. Therefore, solid adsorbents are needed to remove iodine. For all future operations where iodine is present, a silver zeolite adsorber is to be used.

  16. Behavior of radioactive iodine and technetium in the spray calcination of high-level waste

    International Nuclear Information System (INIS)

    Knox, C.A.; Farnsworth, R.K.

    1981-08-01

    The Remote Laboratory-Scale Waste Treatment Facility (RLSWTF) was designed and built as a part of the High-Level Waste Immobilization Program (now the High-Level Waste Process Development Program) at the Pacific Northwest Laboratory. In this facility, which is installed in a radiochemical cell, small volumes of radioactive liquid wastes can be solidified, the process off gas can be analyzed, and the methods for decontaminating this off gas can be tested. Initial operations were completed with nonradioactive, simulated waste solutions (Knox, Siemens and Berger 1981). The first radioactive operations in this facility were performed with a simulated, commercial waste composition containing tracer levels of 99 Tc and 131 I. This report describes the facility and test operations and presents the results of the behavior of 131 I and 99 Tc during solidification of radioactive liquid wastes. During the spray calcination of commercial high-level liquid waste spiked with 99 Tc and 131 I, there was a 0.3 wt% loss of particulates, a 0.15 wt% loss of 99 Tc and a 31 wt% loss of 131 I past the sintered-metal filters. These filters and a venturi scrubber were very efficient in removing particulates and 99 Tc from the off-gas stream. Liquid scrubbers were not efficient in removing 131 I, as 25% of the total lost went to the building off-gas system. Therefore, solid adsorbents will be needed to remove iodine. For all future RLSWTF operations where iodine is present, a silver zeolite adsorber will be used

  17. Vitrification of radioactive high-level waste by spray calcination and in-can melting

    Science.gov (United States)

    Hanson, M. S.; Bjorklund, W. J.

    1980-07-01

    After several nonradioactive test runs, radioactive waste from the processing of 1.5 t of spent, light water reactor fuel was successfully concentrated, dried and converted to a vitreous product. A total of 97 L of waste glass (in two stainless steel canisters) was produced. The spray calcination process coupled to the in-can melting process, as developed at Pacific Northwest Labortory, was used to vitrify the waste. An effluent system consisting of a variety of condensation of scrubbing steps more than adequately decontaminated the process off gas before it was released to the atmosphere.

  18. Technology status of spray calcination--vitrification of high-level liquid waste for full-scale application

    International Nuclear Information System (INIS)

    Keeley, R.B.; Bonner, W.F.; Larson, D.E.

    1977-01-01

    Spray calcination and vitrification technology for stabilization of high-level nuclear wastes has been developed to the point that initiation of technology transfer to an industrial-sized facility could begin. This report discusses current process and equipment development status together with additional R and D studies and engineering evaluations needed. Preliminary full-scale process and equipment descriptions are presented. Technology application in a full-scale plant would blend three distinct maintenance design philosophies, depending on service life anticipated: (1) totally remote maintenance with limited viewing and handling equipment, (2) totally remote maintenance with extensive viewing and handling equipment, and (3) contact maintenance

  19. Design and performance of a full-scale spray calciner for nonradioactive high-level-waste-vitrification studies

    International Nuclear Information System (INIS)

    Miller, F.A.

    1981-06-01

    In the spray calcination process, liquid waste is spray-dried in a heated-wall spray dryer (termed a spray calciner), and then it may be combined in solid form with a glass-forming frit. This mixture is then melted in a continuous ceramic melter or in an in-can melter. Several sizes of spray calciners have been tested at PNL- laboratory scale, pilot scale and full scale. Summarized here is the experience gained during the operation of PNL's full-scale spray calciner, which has solidified approx. 38,000 L of simulated acid wastes and approx. 352,000 L of simulated neutralized wastes in 1830 h of processing time. Operating principles, operating experience, design aspects, and system descriptions of a full-scale spray calciner are discussed. Individual test run summaries are given in Appendix A. Appendices B and C are studies made by Bechtel Inc., under contract by PNL. These studies concern, respectively, feed systems for the spray calciner process and a spray calciner vibration analysis. Appendix D is a detailed structural analysis made at PNL of the spray calciner. These appendices are included in the report to provide a complete description of the spray calciner and to include all major studies made concerning PNL's full-scale spray calciner

  20. Effect of aluminum and silicon reactants and process parameters on glass-ceramic waste form characteristics for immobilization of high-level fluorinel-sodium calcined waste

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1993-06-01

    In this report, the effects of aluminum and silicon reactants, process soak time and the initial calcine particle size on glass-ceramic waste form characteristics for immobilization of the high-level fluorinel-sodium calcined waste stored at the Idaho Chemical Processing Plant (ICPP) are investigated. The waste form characteristics include density, total and normalized elemental leach rates, and microstructure. Glass-ceramic waste forms were prepared by hot isostatically pressing (HIPing) a pre-compacted mixture of pilot plant fluorinel-sodium calcine, Al, and Si metal powders at 1050 degrees C, 20,000 psi for 4 hours. One of the formulations with 2 wt % Al was HIPed for 4, 8, 16 and 24 hours at the same temperature and pressure. The calcine particle size range include as calcined particle size smaller than 600 μm (finer than -30 mesh, or 215 μm Mass Median Diameter, MMD) and 180 μm (finer than 80 mesh, or 49 μm MMD)

  1. Spray calcination of nuclear wastes

    International Nuclear Information System (INIS)

    Bonner, W.F.; Blair, H.T.; Romero, L.S.

    1976-01-01

    The spray calciner is a relatively simple machine; operation is simple and is easily automated. Startup and shutdown can be performed in less than an hour. A wide variety of waste compositions and concentrations can be calcined under easily maintainable conditions. Spray calcination of all commercial fuel reprocessor high-level liquid wastes and mixed high and intermediate-level wastes have been demonstrated. Wastes have been calcined containing over 2M sodium. Thus waste generated during plant startup and shutdown can be blended with normal waste and calcined. Spray calcination of ILLW has also been demonstrated. A remotely replaceable atomizing nozzle has been developed for use in plant scale equipment. The 6 mm (0.25 inch) orifice and ceramic tip offer freedom from plugging and erosion thus nozzle replacement should be required only after several months operation. Calciner capacity of over 75 l/h (20 gal/h) has been demonstrated in pilot scale equipment. Sintered stainless steel filters are effective in deentraining over 99.9 percent of the solids that result from calcining the feedstock. Since such a small amount of radionuclides escape the calciner the volume of recycle required from the effluent treatment system is very small. The noncondensable off-gas volume is also low, less than 0.5 m 3 /min (15 scfm) for a liquid feedrate of 75 l/hr (20 gal/hr). Calcine holdup in the calciner is less than 1 kg, thus the liquid feedrate is directly relatable to calcine flowrate. The calcine produced is very fine and reactive. Successful remote operation and maintenance of a heated wall spray calciner has been demonstrated while processing actual high-level waste. During these operations radionuclide volatilization from the calciner was acceptably low. 8 figures

  2. Comparison of the rotary calciner-metallic melter and the slurry-fed ceramic melter technologies for vitrifying West Valley high-level wastes

    International Nuclear Information System (INIS)

    Chapman, C.C.

    1983-01-01

    Two processes which are believed applicable and available for vitrification of West Valley's high-level (HLW) wastes were technically evaluated and compared. The rotary calciner-metallic melter (AVH) and the slurry-fed ceramic melter (SFCM) were evaluated under the following general categories: process flow sheet, remote operability, safety and environmental considerations, and estimated cost and schedules

  3. Generalized Test Plan for the Vitrification of Simulated High-Level -Waste Calcine in the Idaho National Laboratory's Bench -Scale Cold Crucible Induction Melter

    International Nuclear Information System (INIS)

    Maio, Vince

    2011-01-01

    This Preliminary Idaho National Laboratory (INL) Test Plan outlines the chronological steps required to initially evaluate the validity of vitrifying INL surrogate (cold) High-Level-Waste (HLW) solid particulate calcine in INL's Cold Crucible Induction Melter (CCIM). Its documentation and publication satisfies interim milestone WP-413-INL-01 of the DOE-EM (via the Office of River Protection) sponsored work package, WP 4.1.3, entitled 'Improved Vitrification' The primary goal of the proposed CCIM testing is to initiate efforts to identify an efficient and effective back-up and risk adverse technology for treating the actual HLW calcine stored at the INL. The calcine's treatment must be completed by 2035 as dictated by a State of Idaho Consent Order. A final report on this surrogate/calcine test in the CCIM will be issued in May 2012-pending next fiscal year funding In particular the plan provides; (1) distinct test objectives, (2) a description of the purpose and scope of planned university contracted pre-screening tests required to optimize the CCIM glass/surrogate calcine formulation, (3) a listing of necessary CCIM equipment modifications and corresponding work control document changes necessary to feed a solid particulate to the CCIM, (4) a description of the class of calcine that will be represented by the surrogate, and (5) a tentative tabulation of the anticipated CCIM testing conditions, testing parameters, sampling requirements and analytical tests. Key FY -11 milestones associated with this CCIM testing effort are also provided. The CCIM test run is scheduled to be conducted in February of 2012 and will involve testing with a surrogate HLW calcine representative of only 13% of the 4,000 m3 of 'hot' calcine residing in 6 INL Bin Sets. The remaining classes of calcine will have to be eventually tested in the CCIM if an operational scale CCIM is to be a feasible option for the actual INL HLW calcine. This remaining calcine's make-up is HLW containing

  4. Method for calcining radioactive wastes

    International Nuclear Information System (INIS)

    Bjorklund, W.J.; McElroy, J.L.; Mendel, J.E.

    1979-01-01

    A method for the preparation of radioactive wastes in a low leachability form involves calcining the radioactive waste on a fluidized bed of glass frit, removing the calcined waste to melter to form a homogeneous melt of the glass and the calcined waste, and then solidifying the melt to encapsulate the radioactive calcine in a glass matrix

  5. Volatility of ruthenium-106, technetium-99, and iodine-129, and the evolution of nitrogen oxide compounds during the calcination of high-level, radioactive nitric acid waste

    International Nuclear Information System (INIS)

    Rimshaw, S.J.; Case, F.N.; Tompkins, J.A.

    1980-02-01

    The nitrate anion is the predominant constituent in all high-level nuclear wastes. Formic acid reacts with the nitrate anion to yield noncondensable, inert gases (N 2 or N 2 O), which can be scrubbed free of 106 Ru, 129 I, and 99 Tc radioactivities prior to elimination from the plant by passing through HEPA filters. Treatment of a high-level authentic radioactive waste with two moles of formic acid per mole of nitrate anion leads to a low RuO 4 volatility of about 0.1%, which can be reduced to an even lower level of 0.007% on adding a 15% excess of formic acid. Without pretreatment of the nitrate waste with formic acid, a high RuO 4 volatility of approx. 35% is observed on calcining a 4.0 N HNO 3 solution in quartz equipment at 350 0 C. The RuO 4 volatility falls to approx. 1.0% on decreasing the initial HNO 3 concentration to 1.0 N or lower. It is postulated that thermal denitration of a highly nitrated ruthenium complex leads to the formation of volatile RuO 4 , while decarboxylation of a ruthenium-formate complex leads to the formation of nonvolatile RuO 2 . Wet scrubbing with water is used to remove RuO 4 from the off-gas stream. In all glass equipment, small amounts of particulate RuO 2 are formed in the gas phase by decomposition of RuO 4 . The 99 Tc volatility was found to vary from 0.2 to 1.4% on calcining HNO 3 and HCOOH (formic acid) solutions over the temperature range of 250 to 600 0 C. These unexpectedly low volatilities of 99 Tc are correlated to the high thermal stability limits of various metal pertechnetates and technetates. Iodine volatilities were high, varying from a low of 30% at 350 0 C to a high of 97% at 650 0 C. It is concluded that with a proper selection of pretreatment and operating conditions the 106 Ru and 99 Tc activities can be retained in the calcined solid with recycle of the wet scrubbing solution

  6. Hot isostatically-pressed aluminosilicate glass-ceramic with natural crystalline analogues for immobilizing the calcined high-level nuclear waste at the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Raman, S.

    1993-12-01

    The additives Si, Al, MgO, P 2 O 5 were mechanically blended with fluorinelsodium calcine in varying proportions. The batches were vacuum sealed in stainless steel canisters and hot isostatically pressed at 20,000 PSI and 1000 C for 4 hours. The resulting suite of glass-ceramic waste forms parallels the natural rocks in microstructural and compositional heterogeneity. Several crystalline phases ar analogous in composition and structure to naturally occurring minerals. Additional crystalline phases are zirconia and Ca-Mg borate. The glasses are enriched in silica and alumina. Approximately 7% calcine elements occur dissolved in this glass and the total glass content in the waste forms averages 20 wt%. The remainder of the calcine elements are partitioned into crystalline phases at 75 wt% calcine waste loading. The waste forms were tested for chemical durability in accordance with the MCC1-test procedure. The leach rates are a function of the relative proportions of additives and calcine, which in turn influence the composition and abundances of the glass and crystalline phases. The DOE leach rate criterion of less than 1 g/m 2 -day is met by all the elements B, Cs and Na are increased by lowering the melt viscosity. This is related to increased crystallization or devitrification with increases in MgO addition. This exploratory work has shown that the increases in waste loading occur by preferred partitioning of the calcine components among crystalline and glass phases. The determination of optimum processing parameters in the form of additive concentration levels, homogeneous blending among the components, and pressure-temperature stabilities of phases must be continued to eliminate undesirable effects of chemical composition, microstructure and glass devitrification

  7. Heat transfer in high-level waste management

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  8. High level nuclear wastes

    International Nuclear Information System (INIS)

    Lopez Perez, B.

    1987-01-01

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

  9. New Waste Calcining Facility (NWCF) Waste Streams

    International Nuclear Information System (INIS)

    K. E. Archibald

    1999-01-01

    This report addresses the issues of conducting debris treatment in the New Waste Calcine Facility (NWCF) decontamination area and the methods currently being used to decontaminate material at the NWCF

  10. Summary of Waste Calcination at INTEC

    Energy Technology Data Exchange (ETDEWEB)

    O' Brien, Barry Henry; Newby, Bill Joe

    2000-10-01

    Fluidized-bed calcination at the Idaho Nuclear Technologies and Engineering Center (INTEC, formally called the Idaho Chemical Processing Plant) has been used to solidify acidic metal nitrate fuel reprocessing and incidental wastes wastes since 1961. A summary of waste calcination in full-scale and pilot plant calciners has been compiled for future reference. It contains feed compositions and operating conditions for all the processing campaigns for the original Waste Calcining Facility (WCF), the New Waste Calcining Facility (NWCF) started up in 1982, and numerous small scale pilot plant tests for various feed types. This summary provides a historical record of calcination at INTEC, and will be useful for evaluating calcinability of future wastes.

  11. New Waste Calciner High Temperature Operation

    International Nuclear Information System (INIS)

    Swenson, M.C.

    2000-01-01

    A new Calciner flowsheet has been developed to process the sodium-bearing waste (SBW) in the INTEC Tank Farm. The new flowsheet increases the normal Calciner operating temperature from 500 C to 600 C. At the elevated temperature, sodium in the waste forms stable aluminates, instead of nitrates that melt at calcining temperatures. From March through May 2000, the new high-temperature flowsheet was tested in the New Waste Calcining Facility (NWCF) Calciner. Specific test criteria for various Calciner systems (feed, fuel, quench, off-gas, etc.) were established to evaluate the long-term operability of the high-temperature flowsheet. This report compares in detail the Calciner process data with the test criteria. The Calciner systems met or exceeded all test criteria. The new flowsheet is a visible, long-term method of calcining SBW. Implementation of the flowsheet will significantly increase the calcining rate of SBW and reduce the amount of calcine produced by reducing the amount of chemical additives to the Calciner. This will help meet the future waste processing milestones and regulatory needs such as emptying the Tank Farm

  12. High-level-waste immobilization

    International Nuclear Information System (INIS)

    Crandall, J.L.

    1982-01-01

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

  13. High-Level Radioactive Waste.

    Science.gov (United States)

    Hayden, Howard C.

    1995-01-01

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

  14. Process for solidifying high-level nuclear waste

    Science.gov (United States)

    Ross, Wayne A.

    1978-01-01

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

  15. Waste Calcining Facility remote inspection report

    International Nuclear Information System (INIS)

    Patterson, M.W.; Ison, W.M.

    1994-08-01

    The purpose of the Waste Calcining Facility (WCF) remote inspections was to evaluate areas in the facility which are difficult to access due to high radiation fields. The areas inspected were the ventilation exhaust duct, waste hold cell, adsorber manifold cell, off-gas cell, calciner cell and calciner vessel. The WCF solidified acidic, high-level mixed waste generated during nuclear fuel reprocessing. Solidification was accomplished through high temperature oxidation and evaporation. Since its shutdown in 1981, the WCFs vessels, piping systems, pumps, off-gas blowers and process cells have remained contaminated. Access to the below-grade areas is limited due to contamination and high radiation fields. Each inspection technique was tested with a mock-up in a radiologically clean area before the equipment was taken to the WCF for the actual inspection. During the inspections, essential information was obtained regarding the cleanliness, structural integrity, in-leakage of ground water, indications of process leaks, indications of corrosion, radiation levels and the general condition of the cells and equipment. In general, the cells contain a great deal of dust and debris, as well as hand tools, piping and miscellaneous equipment. Although the building appears to be structurally sound, the paint is peeling to some degree in all of the cells. Cracking and spalling of the concrete walls is evident in every cell, although the east wall of the off-gas cell is the worst. The results of the completed inspections and lessons learned will be used to plan future activities for stabilization and deactivation of the facility. Remote clean-up of loose piping, hand tools, and miscellaneous debris can start immediately while information from the inspections is factored into the conceptual design for deactivating the facility

  16. High Level Radioactive Waste Management

    International Nuclear Information System (INIS)

    1991-01-01

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

  17. Calcine Waste Storage at the Idaho Nuclear Technology and Engineering Center

    Energy Technology Data Exchange (ETDEWEB)

    M. D. Staiger

    1999-06-01

    A potential option in the program for long-term management of high-level wastes at the Idaho Nuclear Technology and Engineering Center (INTEC), at the Idaho National Engineering and Environmental Laboratory, calls for retrieving calcine waste and converting it to a more stable and less dispersible form. An inventory of calcine produced during the period December 1963 to May 1999 has been prepared based on calciner run, solids storage facilities operating, and miscellaneous operational information, which gives the range of chemical compositions of calcine waste stored at INTEC. Information researched includes calciner startup data, waste solution analyses and volumes calcined, calciner operating schedules, solids storage bin capacities, calcine storage bin distributor systems, and solids storage bin design and temperature monitoring records. Unique information on calcine solids storage facilities design of potential interest to remote retrieval operators is given.

  18. High-level radioactive wastes

    International Nuclear Information System (INIS)

    Grissom, M.C.

    1982-10-01

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

  19. High-Level Waste Vitrification Facility Feasibility Study

    International Nuclear Information System (INIS)

    D. A. Lopez

    1999-01-01

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

  20. High-Level Waste Vitrification Facility Feasibility Study

    Energy Technology Data Exchange (ETDEWEB)

    D. A. Lopez

    1999-08-01

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

  1. Optimizing High Level Waste Disposal

    International Nuclear Information System (INIS)

    Dirk Gombert

    2005-01-01

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

  2. Remote ignitability analysis of high-level radioactive waste

    International Nuclear Information System (INIS)

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

    1992-09-01

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

  3. Other-than-high-level waste

    International Nuclear Information System (INIS)

    Bray, G.R.

    1976-01-01

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

  4. The management of high-level radioactive wastes

    International Nuclear Information System (INIS)

    Lennemann, Wm.L.

    1979-01-01

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

  5. Pilot-plant development of a Rover waste calcination flowsheet

    International Nuclear Information System (INIS)

    Birrer, S.A.

    1978-04-01

    Results of eight runs, six using the 10-cm dia and two using the 30-cm dia pilot-plant calciners, in which simulated first-cycle Rover waste was calcined, are described. Results of the tests showed that a feed blend consisting of one volume simulated first-cycle Rover waste and one or two volumes simulated first-cycle zirconium waste could not be successfully calcined. 5 figs., 8 tables

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

    International Nuclear Information System (INIS)

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

    1980-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Rusin, J.M.

    1980-12-01

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

  8. Review of high-level waste form properties

    International Nuclear Information System (INIS)

    Rusin, J.M.

    1980-12-01

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

  9. Design criteria for the new waste calcining facility at the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Anderson, F.H.; Bingham, G.E.; Buckham, J.A.; Dickey, B.R.; Slansky, C.M.; Wheeler, B.R.

    1976-01-01

    The New Waste Calcining Facility (NWCF) at the Idaho Chemical Processing Plant (ICPP) is being built to replace the existing fluidized-bed, high-level waste calcining facility (WCF). Performance of the WCF is reviewed, equipment failures in WCF operation are examined, and pilot-plant studies on calciner improvements are given in relation to NWCF design. Design features of the NWCF are given with emphasis on process and equipment improvements. A major feature of the NWCF is the use of remote maintenance facilities for equipment with high maintenance requirements, thereby reducing personnel exposures during maintenance and reducing downtime resulting from plant decontamination. The NWCF will have a design net processing rate of 11.36 m 3 of high-level waste per day, and will incorporate in-bed combustion of kerosene for heating the fluidized bed calciner. The off-gas cleaning system will be similar to that for the WCF

  10. Current high-level waste solidification technology

    International Nuclear Information System (INIS)

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

    1976-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  12. Application of SYNROC to high-level defense wastes

    International Nuclear Information System (INIS)

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

    1981-01-01

    The SYNROC method for immobilization of high-level nuclear reactor wastes is currently being applied to US defense wastes in tank storage at Savannah River, South Carolina. The minerals zirconolite, perovskite, and hollandite are used in SYNROC D formulations to immobilize fission products and actinides that comprise up to 10% of defense waste sludges and coexisting solutions. Additional phase in SYNROC D are nepheline, the host phase for sodium; and spinel, the host for excess aluminum and iron. Up to 70 wt % of calcined sludge can be incorporated with 30 wt % of SYNROC additives to produce a waste form consisting of 10% nepheline, 30% spinel, and approximately 20% each of the radioactive waste-bearing phases. Urea coprecipitation and spray drying/calcining methods have been used in the laboratory to produce homogeneous, reactive ceramic powders. Hot pressing and sintering at temperatures from 1000 to 1100 0 C result in waste form products with greater than 97% of theoretical density. Hot isostatic pressing has recently been implemented as a processing alternative. Characterization of waste-form mineralogy has been done by means of XRD, SEM, and electron microprobe. Leaching of SYNROC D samples is currently being carried out. Assessment of radiation damage effects and physical properties of SYNROC D will commence in FY81

  13. Remotely replaceable fuel and feed nozzles for the new waste calcining facility calciner vessel

    International Nuclear Information System (INIS)

    Fletcher, R.D.; Carter, J.A.; May, K.W.

    1978-01-01

    The development and testing of remotely replaceable fuel and feed nozzles for calcination of liquid radioactive wastes in the calciner vessel of the New Waste Calcining Facility being built at the Idaho National Engineering Laboratory is described. A complete fuel nozzle assembly was fabricated and tested at the Remote Maintenance Development Facility to evolve design refinements, identify required support equipment, and develop handling techniques. The design also provided for remote replacement of the nozzle support carriage and adjacent feed and fuel pipe loops using two pairs of master-slave manipulators

  14. Formulation Efforts for Direct Vitrification of INEEL Blend Calcine Waste Simulate: Fiscal Year 2000

    Energy Technology Data Exchange (ETDEWEB)

    Crum, Jarrod V.; Vienna, John D.; Peeler, David K.; Reamer, I. A.

    2001-03-30

    This report documents the results of glass formulation efforts for Idaho National Engineering and Environmental Laboratory (INEEL) high level waste (HWL) calcine. Two waste compositions were used during testing. Testing started by using the Run 78 calcine composition and switched to simulated Blend calcine composition when it became available. The goal of the glass formulation efforts was to develop a frit composition that will accept higher waste loading that satisfies the glass processing and product acceptance constraints. 1. Melting temperature of 1125 ? 25?C 2. Viscosity between 2 and 10 Pa?s at the melting temperature 3. Liquidus temperature at least 100?C below the melting temperature 4. Normalized release of B, Li and Na each below 1 g/m2 (per ASTM C 1285-97) Glass formulation efforts tested several frit compositions with variable waste loadings of Run 78 calcine waste simulant. Frit 107 was selected as the primary candidate for processing since it met all process and performance criteria up to 45 mass% waste loading. When the simulated Blend calcine waste composition became available Frits 107 and 108 compositions were retested and again Frit 107 remained the primary candidate. However, both frits suffered a decrease in waste loading when switching from the Run 78 calcine to simulated Blend calcine waste composition. This was due to increase concentrations of both F and Al2O3 along with a decrease in CaO and Na2O in the simulate Blend calcine waste all of which have strong impacts on the glass properties that limit waste loading of this type of waste.

  15. Design and development of a rotary calciner for radiochemical waste

    International Nuclear Information System (INIS)

    Pande, D.P.; Sutar, V.D.; Sengar, P.B.S.

    1997-01-01

    Present experience and knowledge in handling of radioactive waste has led to identification of major thrust areas in the development of the treatment processes. In order to reduce evaporation and volatility losses in the vitrification facility, it is advantageous to carry out evaporation and calcination steps in another equipment like rotary calciner. Efforts have been directed for the engineering development of a Rotary Ball Kiln calciner. This paper highlights the important design features of the Rotary Ball Kiln Calciner for the radioactive waste. In this work, an attempt has been made to systematically evaluate the influence of process and design parameters. The results obtained on calcination will provide a design basis and rational methodology for the optimum utilization of these processes and equipment for volume reduction and calcination of the liquid waste

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  17. PAIRWISE BLENDING OF HIGH LEVEL WASTE

    International Nuclear Information System (INIS)

    CERTA, P.J.

    2006-01-01

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

  18. Determination of the Rate of Formation of Hydroceramic Waste Forms made with INEEL Calcined Wastes; FINAL

    International Nuclear Information System (INIS)

    Barry Scheetz; Johnson Olanrewaju

    2001-01-01

    The formulation, synthesis, characterization and hydration kinetics of hydroceramic waste forms designed as potential hosts for existing INEEL calcine high-level wastes have been established as functions of temperature and processing time. Initial experimentations were conducted with several aluminosilicate pozzolanic materials, ranging from fly ash obtained from various power generating coal and other combustion industries to reactive alumina, natural clays and ground bottled glass powders. The final selection criteria were based on the ease of processing, excellent physical properties and chemical durability (low-leaching) determined from the PCT test produced in hydroceramic. The formulation contains vermiculite, Sr(NO32), CsC1, NaOH, thermally altered (calcined natural clay) and INEEL simulated calcine high-level nuclear wastes and 30 weight percent of fluorinel blend calcine and zirconia calcine. Syntheses were carried out at 75-200 degree C at autogeneous water pressure (100% relative humidity) at various time intervals. The resulting monolithic compact products were hard and resisted breaking when dropped from a 5 ft height. Hydroceramic host mixed with fluorinel blend calcine and processed at 75 degree C crumbled into rice hull-side grains or developed scaly flakes. However, the samples equally possessed the same chemical durability as their unbroken counterparts. Phase identification by XRD revealed that hydroceramic host crystallized type zeolite at 75-150 degree C and NaP1 at 175-200 degree C in addition to the presence of quartz phase originating from the clay reactant. Hydroceramic host mixed with either fluorinel blend calcine or zirconia calcine crystallized type A zeolite at 75-95 degree C, formed a mixture of type A zeolite and hydroxysodalite at 125-150 degree C and hydroxysodalite at 175-200 degree C. Quartz, calcium fluoride and zirconia phases from the clay reactant and the two calcine wastes were also detected. The PCT test solution

  19. Determination of the Rate of Formation of Hydroceramic Waste Forms made with INEEL Calcined Wastes

    Energy Technology Data Exchange (ETDEWEB)

    Barry Scheetz; Johnson Olanrewaju

    2001-10-15

    The formulation, synthesis, characterization and hydration kinetics of hydroceramic waste forms designed as potential hosts for existing INEEL calcine high-level wastes have been established as functions of temperature and processing time. Initial experimentations were conducted with several aluminosilicate pozzolanic materials, ranging from fly ash obtained from various power generating coal and other combustion industries to reactive alumina, natural clays and ground bottled glass powders. The final selection criteria were based on the ease of processing, excellent physical properties and chemical durability (low-leaching) determined from the PCT test produced in hydroceramic. The formulation contains vermiculite, Sr(NO32), CsC1, NaOH, thermally altered (calcined natural clay) and INEEL simulated calcine high-level nuclear wastes and 30 weight percent of fluorinel blend calcine and zirconia calcine. Syntheses were carried out at 75-200 degree C at autogeneous water pressure (100% relative humidity) at various time intervals. The resulting monolithic compact products were hard and resisted breaking when dropped from a 5 ft height. Hydroceramic host mixed with fluorinel blend calcine and processed at 75 degree C crumbled into rice hull-side grains or developed scaly flakes. However, the samples equally possessed the same chemical durability as their unbroken counterparts. Phase identification by XRD revealed that hydroceramic host crystallized type zeolite at 75-150 degree C and NaP1 at 175-200 degree C in addition to the presence of quartz phase originating from the clay reactant. Hydroceramic host mixed with either fluorinel blend calcine or zirconia calcine crystallized type A zeolite at 75-95 degree C, formed a mixture of type A zeolite and hydroxysodalite at 125-150 degree C and hydroxysodalite at 175-200 degree C. Quartz, calcium fluoride and zirconia phases from the clay reactant and the two calcine wastes were also detected. The PCT test solution

  20. Cermets for high level waste containment

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  1. Final report, Task 2: alternative waste management options, Nuclear Fuel Services, Inc., high level waste

    International Nuclear Information System (INIS)

    1978-01-01

    Of the alternatives considered for disposal of the high-level waste in tanks 8D2 and 8D4, the following process is recommended: homogenization of the contents of tank 8D2, centrifugation of the sludge and supernate, mixing of the 8D4 acid waste with the centrifuged sludge, and converting the mixture to a borosilicate glass using the Hanford spray calciner/in-can melter

  2. Calcination/dissolution testing for Hanford Site tank wastes

    International Nuclear Information System (INIS)

    Colby, S.A.; Delegard, C.H.; McLaughlin, D.F.; Danielson, M.J.

    1994-07-01

    Thermal treatment by calcination offers several benefits for the treatment of Hanford Site tank wastes, including the destruction of organics and ferrocyanides and an hydroxide fusion that permits the bulk of the mostly soluble nonradioactive constituents to be easily separated from the insoluble transuranic residue. Critical design parameters were tested, including: (1) calciner equipment design, (2) hydroxide fusion chemistry, and (3) equipment corrosion. A 2 gal/minute pilot plant processed a simulated Tank 101-SY waste and produced a free flowing 700 C molten calcine with an average calciner retention time of 20 minutes and >95% organic, nitrate, and nitrite destruction. Laboratory experiments using actual radioactive tank waste and the simulated waste pilot experiments indicate that 98 wt% of the calcine produced is soluble in water, leaving an insoluble transuranic fraction. All of the Hanford Site tank wastes can benefit from calcination/dissolution processing, contingent upon blending various tank waste types to ensure a target of 70 wt% sodium hydroxide/nitrate/nitrite fluxing agent. Finally, corrosion testing indicates that a jacketed nickel liner cooled to below 400 C would corrode <2 mil/year (0.05 mm/year) from molten calcine attack

  3. Chemistry of proposed calcination/dissolution processing of Hanford Site tank wastes

    International Nuclear Information System (INIS)

    Delegard, C.H.

    1995-01-01

    Plans exist to separate radioactive waste stored in underground tanks at the US Department of Energy's Hanford Site in south central Washington State into low-level and high-level fractions, and to immobilize the separate fractions in high-integrity vitrified forms for long-term disposal. Calcination with water dissolution has been proposed as a possible treatment for achieving low/high-level separation. Chemistry development activities conducted since 1992 with simulated and genuine tank waste show that calcination/dissolution destroys organic carbon and converts nitrate and nitrite to hydroxide and benign offgases. The process also dissolves significant quantities of bulk chemicals (aluminum, chromium, and phosphate), allowing their redistribution from the high-level to the low-level fraction. Present studies of the chemistry of calcination/dissolution processing of genuine wastes, conducted in the period October 1993 to September 1994, show the importance of sodium fluoride phosphate double salt in controlling phosphate dissolution. Peptization of waste solids is of concern if extensive washing occurs. Strongly oxidizing conditions imposed by calcination reactions were found to convert transition metals to soluble anions in the order chromate > manganate > > ferrate. In analogy with manganese behavior, plutonium dissolution, presumably by oxidation to more soluble anionic species, also occurs by calcination/dissolution. Methods to remove plutonium from the product low-level solution stream must be developed

  4. High-Level Waste Melter Study Report

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-13

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

  5. High-level waste processing and disposal

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  6. High-level radioactive wastes. Supplement 1

    International Nuclear Information System (INIS)

    McLaren, L.H.

    1984-09-01

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

  7. Materials for high-level waste containment

    International Nuclear Information System (INIS)

    Marsh, G.P.

    1982-01-01

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

  8. Immobilisation of high level nuclear reactor wastes in SYNROC

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-03-15

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

  9. Timing of High-level Waste Disposal

    International Nuclear Information System (INIS)

    2008-01-01

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

  10. Disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    Glasby, G.P.

    1977-01-01

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

  11. Powder technological vitrification of simulated high-level waste

    International Nuclear Information System (INIS)

    Gahlert, S.

    1988-03-01

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

  12. International high-level radioactive waste repositories

    International Nuclear Information System (INIS)

    Lin, W.

    1996-01-01

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

  13. Ramifications of defining high-level waste

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  14. Technetium Chemistry in High-Level Waste

    International Nuclear Information System (INIS)

    Hess, Nancy J.

    2006-01-01

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

  15. Fixation of calcined waste by bituminization or cementation

    International Nuclear Information System (INIS)

    Napravnik, J.; Kyrs, M.; Ditl, P.

    1983-01-01

    The overall concept is given of the combination of calcination with fixation into bitumen, cement etc. The design is shown of a calciner with the capacity of 10 L/h which was tested on real radioactive wastes for 2000 h. The geometrical and operating parameters of the apparatus have been optimized based on a statistical evaluation of the experiments. Wastes containing nitrates are calcined at 300-550 deg. C, yielding oxides. Wastes containing sulphates, carbonates, KMnO 4 , or borates are calcined at 150-330 deg. C, yielding soluble salts. The content of H 3 BO 4 and Na 2 B 4 O 7 and in some cases of sulphates in the calcinate retards hardening of the mixture with cement. Nitrates and detergents also interfere. The effect of the above components on the products mixed with bitumen is much less. Detergents can be decomposed at 200-300 deg. C; organic acids can be reacted with A1 salts to form insoluble substances lowering the leaching rate of Sr and Cs; small amounts of SiO 2 eliminate the effect of borates on cement hardening. The drawbacks of bituminization with bitumen emulsions are the complicated preparation of the emulsion, higher leaching rate of the product and low stability of the emulsion against breaking. The leachability was determined (1-50 days) of different products containing LWR wastes: 33% of concentrated waste in cement of calcination product stabilized with PVA exhibit approx. 8x10 - 3 g/cm 2 per day, 33% of calcine in cement approx. 3x10 - 3 ; 40% concentrate fixed with bitumen emulsion approx. 9x10 - 4 ; 50% calcine stabilized with PVA in bitumen, pilot-plant scale approx. 2x10 - 5 ; the same but on a laboratory scale approx. 1.10 - 5 . (author)

  16. High-level radioactive wastes. Supplement 1

    Energy Technology Data Exchange (ETDEWEB)

    McLaren, L.H. (ed.)

    1984-09-01

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

  17. Decommissioning high-level waste surface facilities

    International Nuclear Information System (INIS)

    1978-04-01

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

  18. Attrition, elutriation, and growth of particles produced in fluidized-bed waste calciners

    International Nuclear Information System (INIS)

    McDonald, F.N.

    1982-09-01

    The Idaho Chemical Processing Plant reduces the volume of high-level liquid radioactive wastes in a fluidized bed to produce a granular calcine product. In the past, difficulties have been experienced in controlling the product's particle size when processing certain blends of sodium-bearing waste. Therefore, experiments in attrition, elutriation, and particle growth were done to characterize how best to control these three parameters. 15 figures, 16 tables

  19. High-level waste processing and disposal

    International Nuclear Information System (INIS)

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

    1984-11-01

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

  20. High level waste fixation in cermet form

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  1. High-level nuclear waste disposal

    International Nuclear Information System (INIS)

    Burkholder, H.C.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Decker, L.A.

    1998-04-01

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

  3. National high-level waste systems analysis

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  4. Management of high level radioactive waste

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  5. Processing vessel for high level radioactive wastes

    International Nuclear Information System (INIS)

    Maekawa, Hiromichi

    1998-01-01

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

  6. Vitrification of high-level liquid wastes

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  7. Ocean disposal of high level radioactive waste

    International Nuclear Information System (INIS)

    1983-01-01

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

  8. Pelleted waste form for high-level ICPP wastes

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  9. Pelleted waste form for high-level ICPP wastes

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  10. High-level radioactive waste management

    International Nuclear Information System (INIS)

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

    1974-01-01

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

  11. Intergenerational ethics of high level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-03-01

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

  12. Intergenerational ethics of high level radioactive waste

    International Nuclear Information System (INIS)

    Takeda, Kunihiko; Nasu, Akiko; Maruyama, Yoshihiro

    2003-01-01

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

  13. Ninth Processing Campaign in the Waste Calcining Facility

    International Nuclear Information System (INIS)

    Childs, K.F.; Donovan, R.I.; Swenson, M.C.

    1982-04-01

    This report discusses the Ninth (and final) Processing Campaign at the Waste Calcining Facility. Several processing interruptions were experienced during this campaign and the emphasis of this report is on process and equipment performance with operating problems and corrective actions discussed in detail

  14. Screening Level Risk Assessment for the New Waste Calcining Facility

    Energy Technology Data Exchange (ETDEWEB)

    M. L. Abbott; K. N. Keck; R. E. Schindler; R. L. VanHorn; N. L. Hampton; M. B. Heiser

    1999-05-01

    This screening level risk assessment evaluates potential adverse human health and ecological impacts resulting from continued operations of the calciner at the New Waste Calcining Facility (NWCF) at the Idaho Nuclear Technology and Engineering Center (INTEC), Idaho National Engineering and Environmental Laboratory (INEEL). The assessment was conducted in accordance with the Environmental Protection Agency (EPA) report, Guidance for Performing Screening Level Risk Analyses at Combustion Facilities Burning Hazardous Waste. This screening guidance is intended to give a conservative estimate of the potential risks to determine whether a more refined assessment is warranted. The NWCF uses a fluidized-bed combustor to solidify (calcine) liquid radioactive mixed waste from the INTEC Tank Farm facility. Calciner off volatilized metal species, trace organic compounds, and low-levels of radionuclides. Conservative stack emission rates were calculated based on maximum waste solution feed samples, conservative assumptions for off gas partitioning of metals and organics, stack gas sampling for mercury, and conservative measurements of contaminant removal (decontamination factors) in the off gas treatment system. Stack emissions were modeled using the ISC3 air dispersion model to predict maximum particulate and vapor air concentrations and ground deposition rates. Results demonstrate that NWCF emissions calculated from best-available process knowledge would result in maximum onsite and offsite health and ecological impacts that are less then EPA-established criteria for operation of a combustion facility.

  15. DEFENSE HIGH LEVEL WASTE GLASS DEGRADATION

    International Nuclear Information System (INIS)

    Ebert, W.

    2001-01-01

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

  16. Design features of a full-scale high-level waste vitrification system

    International Nuclear Information System (INIS)

    Siemens, D.H.; Bonner, W.F.

    1976-08-01

    A system has been designed and is currently under construction for vitrification of commercial high-level waste. The process consists of a spray calciner coupled to an in-can melter. Due to the high radiation levels expected, this equipment is designed for totally remote operation and maintenance. The in-cell arrangement of this equipment has been developed cooperatively with a nuclear fuel reprocessor. The system will be demonstrated both full scale with nonradioactive simulated waste and pilot scale with actual high-level waste

  17. Vitrification of high level wastes in France

    International Nuclear Information System (INIS)

    Sombret, C.

    1984-02-01

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

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

    International Nuclear Information System (INIS)

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

    1978-05-01

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

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

    International Nuclear Information System (INIS)

    Fehringer, D.J.

    1985-10-01

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

  20. 40 CFR 227.30 - High-level radioactive waste.

    Science.gov (United States)

    2010-07-01

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

  1. Durability, mechanical, and thermal properties of experimental glass-ceramic forms for immobilizing ICPP high level waste

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1990-01-01

    The high-level liquid waste generated at the Idaho Chemical Processing Plant (ICPP) is routinely solidified into granular calcined high-level waste (HLW) and stored onsite. Research is being conducted at the ICPP on methods of immobilizing the HLW, including developing a durable glass-ceramic form which has the potential to significantly reduce the final waste volume by up to 60% compared to a glass form. Simulated, pilot plant, non-radioactive, calcines similar to the composition of the calcined HLW and glass forming additives are used to produce experimental glass-ceramic forms. The objective of the research reported in this paper is to study the impact of ground calcine particle size on durability and mechanical and thermal properties of experimental glass-ceramic forms

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

    International Nuclear Information System (INIS)

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

    1995-09-01

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

  3. Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center

    Energy Technology Data Exchange (ETDEWEB)

    Staiger, M. Daniel, Swenson, Michael C.

    2011-09-01

    This comprehensive report provides definitive volume, mass, and composition (chemical and radioactivity) of calcined waste stored at the Idaho Nuclear Technology and Engineering Center. Calcine composition data are required for regulatory compliance (such as permitting and waste disposal), future treatment of the caline, and shipping the calcine to an off-Site-facility (such as a geologic repository). This report also contains a description of the calcine storage bins. The Calcined Solids Storage Facilities (CSSFs) were designed by different architectural engineering firms and built at different times. Each CSSF has a unique design, reflecting varying design criteria and lessons learned from historical CSSF operation. The varying CSSF design will affect future calcine retrieval processes and equipment. Revision 4 of this report presents refinements and enhancements of calculations concerning the composition, volume, mass, chemical content, and radioactivity of calcined waste produced and stored within the CSSFs. The historical calcine samples are insufficient in number and scope of analysis to fully characterize the entire inventory of calcine in the CSSFs. Sample data exist for all the liquid wastes that were calcined. This report provides calcine composition data based on liquid waste sample analyses, volume of liquid waste calcined, calciner operating data, and CSSF operating data using several large Microsoft Excel (Microsoft 2003) databases and spreadsheets that are collectively called the Historical Processing Model. The calcine composition determined by this method compares favorably with historical calcine sample data.

  4. Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center

    International Nuclear Information System (INIS)

    Staiger, M. Daniel; Swenson, Michael C.

    2011-01-01

    This comprehensive report provides definitive volume, mass, and composition (chemical and radioactivity) of calcined waste stored at the Idaho Nuclear Technology and Engineering Center. Calcine composition data are required for regulatory compliance (such as permitting and waste disposal), future treatment of the caline, and shipping the calcine to an off-Site-facility (such as a geologic repository). This report also contains a description of the calcine storage bins. The Calcined Solids Storage Facilities (CSSFs) were designed by different architectural engineering firms and built at different times. Each CSSF has a unique design, reflecting varying design criteria and lessons learned from historical CSSF operation. The varying CSSF design will affect future calcine retrieval processes and equipment. Revision 4 of this report presents refinements and enhancements of calculations concerning the composition, volume, mass, chemical content, and radioactivity of calcined waste produced and stored within the CSSFs. The historical calcine samples are insufficient in number and scope of analysis to fully characterize the entire inventory of calcine in the CSSFs. Sample data exist for all the liquid wastes that were calcined. This report provides calcine composition data based on liquid waste sample analyses, volume of liquid waste calcined, calciner operating data, and CSSF operating data using several large Microsoft Excel (Microsoft 2003) databases and spreadsheets that are collectively called the Historical Processing Model. The calcine composition determined by this method compares favorably with historical calcine sample data.

  5. Volatilization behavior of semivolatile elements in vitrification of high-level liquid waste

    International Nuclear Information System (INIS)

    Igarashi, Hiroshi; Kato, Koh; Takahashi, Takeshi

    1991-11-01

    The effect of temperature on the volatilization of ruthenium, technetium, and selenium was observed in calcination experiments with simulated high-level liquid waste. Technetium and selenium were more volatile as calcining temperature increased. Ruthenium was less volatile when temperature exceeded 300degC. More than 80% of ruthenium that volatilized from room temperature to 500degC occurred between 200 and 300degC. A small amount of ruthenium volatilized above 300degC as well as below 135degC. (author)

  6. 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

  7. Idaho National Engineering Laboratory High-Level Waste Roadmap

    International Nuclear Information System (INIS)

    1993-08-01

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

  8. Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center

    Energy Technology Data Exchange (ETDEWEB)

    M. D. Staiger

    2007-06-01

    This report provides a quantitative inventory and composition (chemical and radioactivity) of calcined waste stored at the Idaho Nuclear Technology and Engineering Center. From December 1963 through May 2000, liquid radioactive wastes generated by spent nuclear fuel reprocessing were converted into a solid, granular form called calcine. This report also contains a description of the calcine storage bins.

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

    International Nuclear Information System (INIS)

    1976-04-01

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

  10. Stability of High-Level Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Besmann, Theodore M.; Vienna, John D.

    2005-09-30

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

  11. In Vitro Studies Evaluating Leaching of Mercury from Mine Waste Calcine Using Simulated Human Body Fluids

    OpenAIRE

    Gray, John E.; Plumlee, Geoffrey S.; Morman, Suzette A.; Higueras, Pablo L.; Crock, James G.; Lowers, Heather A.; Witten, Mark L.

    2010-01-01

    In vitro bioaccessibility (IVBA) studies were carried out on samples of mercury (Hg) mine-waste calcine (roasted Hg ore) by leaching with simulated human body fluids. The objective was to estimate potential human exposure to Hg due to inhalation of airborne calcine particulates and hand-to-mouth ingestion of Hg-bearing calcines. Mine waste calcines collected from Hg mines at Almad?n, Spain, and Terlingua, Texas, contain Hg sulfide, elemental Hg, and soluble Hg compounds, which constitute prim...

  12. Disposal of high-level radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-03-01

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

  13. High level waste management in France

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  14. Volatilities of ruthenium, iodine, and technetium on calcining fission product nitrate wastes

    International Nuclear Information System (INIS)

    Rimshaw, S.J.; Case, F.N.

    1980-01-01

    Various high-level nitrate wastes were subjected to formic acid denitration. Formic acid reacts with the nitrate anion to yield noncondensable, inert gases according to the following equation: 4 HCOOH + 2 HNO 3 → N 2 O + 4 CO 2 + 5 H 2 O. These gases can be scrubbed free of 106 Ru, 131 I, and 99 Tc radioactivities prior to elimination from the plant by passage through HEPA filters. The formation of deleterious NO/sub x/ is avoided. Moreover, formic acid reduces ruthenium to a lower valence state with a sharp reduction in RuO 4 volatility during subsequent calcination of the pretreated waste. It is shown that a minimum of 3% of RuO 4 in an off-gas stream reacts with Davison silica gel (Grade 40) to give a fine RuO 2 aerosol having a particle size of 0.5 μ. This RuO 2 aerosol passes through water or weak acid scrub solutions but is trapped by a caustic scrub solution. Iodine volatilizes almost completely on calcining an acidic waste, and the iodine volatility increases with increasing calcination temperature. On calcining an alkaline sodium nitrate waste the iodine volatility is about an order of magnitude lower, with a relatively low iodine volatility of 0.39% at a calcination temperature of 250 0 C and a moderate volatility of 9.5% at 600 0 C. Volatilities of 99 Tc were generally 0 C. Data are presented to indicate that 99 Tc concentrates in the alkaline sodium nitrate supernatant waste, with approx. 10 mg 99 Tc being associated with each curie of 137 Cs present in the waste. It is shown that lutidine (2,4 dimethyl-pyridine) extracts Tc(VII) quantitatively from alkaline supernatant wastes. The distribution coefficient (K/sub D/) for Tc(VII) going into the organic phase in the above system is 102 for a simulated West Valley waste and 191 for a simulated Savannah River Plant (SRP) waste

  15. Glasses used for the high level radioactive wastes storage

    International Nuclear Information System (INIS)

    Sombret, C.

    1983-06-01

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

  16. High-level waste immobilization program: an overview

    International Nuclear Information System (INIS)

    Bonner, W.R.

    1979-09-01

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

  17. Multipurpose optimization models for high level waste vitrification

    International Nuclear Information System (INIS)

    Hoza, M.

    1994-08-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  20. Remote process connectors for the new waste calcining facility

    International Nuclear Information System (INIS)

    Jacobs, R.T.; Carter, J.A.; Hohback, A.C.

    1978-01-01

    The remote process connectors developed, used, and tested at the Remote Maintenance Development Facility are described. These connectors, including the three-bolt kinematic-graphite flange and watertight electrical connectors, are assembled on master jigs (holding-welding fixture) to form interchangeable pump and valve loop assemblies. These assemblies, with their guide-in platforms, make possible a method of performing remote maintenance at the New Waste Calcining Facility which is a departure from methods that until now have been the standard of the industry

  1. Long-term high-level waste technology program

    International Nuclear Information System (INIS)

    1980-04-01

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

  2. Characteristics of solidified high-level waste products

    International Nuclear Information System (INIS)

    1979-01-01

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

  3. Criteria for high-level waste disposal

    International Nuclear Information System (INIS)

    Sousselier, Y.

    1981-01-01

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

  4. Vitrification of high-level alumina nuclear waste

    International Nuclear Information System (INIS)

    Brotzman, J.R.

    1979-01-01

    Borophosphate glass compositions have been developed for the vitrification of a high-alumina calcined defense waste. The effect of substituting SiO 2 , P 2 O 5 and CuO for B 2 O 3 on the viscosity and leach resistance was measured. The effect of the alkali to borate ratio and the Li 2 O:Na 2 O ratio on the melt viscosity and leach resistance was also measured

  5. Chemistry of application of calcination/dissolution to the Hanford tank waste inventory

    International Nuclear Information System (INIS)

    Delegard, C.H.; Elcan, T.D.; Hey, B.E.

    1994-05-01

    Approximately 330,000 metric tons of sodium-rich radioactive waste originating from separation of plutonium from irradiated uranium fuel are stored in underground tanks at the Hanford Site in Washington State. Fractionation of the waste into low-level waste (LLW) and high-level waste (HLW) streams is envisioned via partial water dissolution and limited radionuclide extraction operations. Under optimum conditions, LLW would contain most of the chemical bulk while HLW would contain virtually all of the transuranic and fission product activity. Calcination at around 850 C, followed by water dissolution, has been proposed as an alternative initial treatment of Hanford Site waste to improve waste dissolution and the envisioned LLW/HLW split. Results of literature and laboratory studies are reported on the application of calcination/dissolution (C/D) to the fractionation of the Hanford Site tank waste inventory. Both simulated and genuine Hanford Site waste materials were used in the lab tests. To evaluation confirmed that C/D processing reduced the amount of several components from the waste. The C/D dissolutions of aluminum and chromium allow redistribution of these waste components from the HLW to the LLW fraction. Comparisons of simple water-washing with C/D processing of genuine Hanford Site waste are also reported based on material (radionuclide and chemical) distributions to solution and solid residue phases. The lab results show that C/D processing yielded superior dissolution of aluminum and chromium sludges compared to simple water dissolution. 57 refs., 26 figs., 18 tabs

  6. Chemistry of application of calcination/dissolution to the Hanford tank waste inventory

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, C.H.; Elcan, T.D.; Hey, B.E.

    1994-05-01

    Approximately 330,000 metric tons of sodium-rich radioactive waste originating from separation of plutonium from irradiated uranium fuel are stored in underground tanks at the Hanford Site in Washington State. Fractionation of the waste into low-level waste (LLW) and high-level waste (HLW) streams is envisioned via partial water dissolution and limited radionuclide extraction operations. Under optimum conditions, LLW would contain most of the chemical bulk while HLW would contain virtually all of the transuranic and fission product activity. Calcination at around 850 C, followed by water dissolution, has been proposed as an alternative initial treatment of Hanford Site waste to improve waste dissolution and the envisioned LLW/HLW split. Results of literature and laboratory studies are reported on the application of calcination/dissolution (C/D) to the fractionation of the Hanford Site tank waste inventory. Both simulated and genuine Hanford Site waste materials were used in the lab tests. To evaluation confirmed that C/D processing reduced the amount of several components from the waste. The C/D dissolutions of aluminum and chromium allow redistribution of these waste components from the HLW to the LLW fraction. Comparisons of simple water-washing with C/D processing of genuine Hanford Site waste are also reported based on material (radionuclide and chemical) distributions to solution and solid residue phases. The lab results show that C/D processing yielded superior dissolution of aluminum and chromium sludges compared to simple water dissolution. 57 refs., 26 figs., 18 tabs.

  7. Sustainable Development and High Level Waste

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  8. Sustainable Development and High Level Waste

    International Nuclear Information System (INIS)

    Jensen, Mikael

    2001-01-01

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

  9. Handling and storage of conditioned high-level wastes

    International Nuclear Information System (INIS)

    1983-01-01

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

  10. Managing commercial high-level radioactive waste

    International Nuclear Information System (INIS)

    1983-01-01

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

  11. Properties and characteristics of high-level waste glass

    International Nuclear Information System (INIS)

    Ross, W.A.

    1977-01-01

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

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

    International Nuclear Information System (INIS)

    Stegen, G.E.

    1996-01-01

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

  13. 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

  14. Overview: Defense high-level waste technology program

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  15. Design and operation of off-gas cleaning systems at high level liquid waste conditioning facilities

    International Nuclear Information System (INIS)

    1988-01-01

    The immobilization of high level liquid wastes from the reprocessing of irradiated nuclear fuels is of great interest and serious efforts are being undertaken to find a satisfactory technical solution. Volatilization of fission product elements during immobilization poses the potential for the release of radioactive substances to the environment and necessitates effective off-gas cleaning systems. This report describes typical off-gas cleaning systems used in the most advanced high level liquid waste immobilization plants and considers most of the equipment and components which can be used for the efficient retention of the aerosols and volatile contaminants. In the case of a nuclear facility consisting of several different facilities, release limits are generally prescribed for the nuclear facility as a whole. Since high level liquid waste conditioning (calcination, vitrification, etc.) facilities are usually located at fuel reprocessing sites (where the majority of the high level liquid wastes originates), the off-gas cleaning system should be designed so that the airborne radioactivity discharge of the whole site, including the emission of the waste conditioning facility, can be kept below the permitted limits. This report deals with the sources and composition of different kinds of high level liquid wastes and describes briefly the main high level liquid waste solidification processes examining the sources and characteristics of the off-gas contaminants to be retained by the off-gas cleaning system. The equipment and components of typical off-gas systems used in the most advanced (large pilot or industrial scale) high level liquid waste solidification plants are described. Safety considerations for the design and safe operation of the off-gas systems are discussed. 60 refs, 31 figs, 17 tabs

  16. Handling and storage of conditioned high-level wastes

    International Nuclear Information System (INIS)

    Heafield, W.

    1984-01-01

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

  17. National high-level waste systems analysis report

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  18. National high-level waste systems analysis report

    International Nuclear Information System (INIS)

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

    1995-09-01

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

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

    International Nuclear Information System (INIS)

    1991-05-01

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

  20. Calcination/dissolution residue treatment

    International Nuclear Information System (INIS)

    Knight, R.C.; Creed, R.F.; Patello, G.K.; Hollenberg, G.W.; Buehler, M.F.; O'Rourke, S.M.; Visnapuu, A.; McLaughlin, D.F.

    1994-09-01

    Currently, high-level wastes are stored underground in steel-lined tanks at the Hanford site. Current plans call for the chemical pretreatment of these wastes before their immobilization in stable glass waste forms. One candidate pretreatment approach, calcination/dissolution, performs an alkaline fusion of the waste and creates a high-level/low-level partition based on the aqueous solubilities of the components of the product calcine. Literature and laboratory studies were conducted with the goal of finding a residue treatment technology that would decrease the quantity of high-level waste glass required following calcination/dissolution waste processing. Four elements, Fe, Ni, Bi, and U, postulated to be present in the high-level residue fraction were identified as being key to the quantity of high-level glass formed. Laboratory tests of the candidate technologies with simulant high-level residues showed reductive roasting followed by carbonyl volatilization to be successful in removing Fe, Ni, and Bi. Subsequent bench-scale tests on residues from calcination/dissolution processing of genuine Hanford Site tank waste showed Fe was separated with radioelement decontamination factors of 70 to 1,000 times with respect to total alpha activity. Thermodynamic analyses of the calcination of five typical Hanford Site tank waste compositions also were performed. The analyses showed sodium hydroxide to be the sole molten component in the waste calcine and emphasized the requirement for waste blending if fluid calcines are to be achieved. Other calcine phases identified in the thermodynamic analysis indicate the significant thermal reconstitution accomplished in calcination

  1. Retrofit design of remotely removable decontamination spray nozzles for the new waste calcining facility at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Gay, J.A.

    1988-01-01

    High level radioactive liquid waste is converted to a solid form at the Idaho Chemical Processing Plant (ICPP). The conversion is done by a fluidized bed combustion process in the calciner vessel. The interior decontamination system for the calciner vessel uses a common header bolted to four decontamination nozzles around the upper head. The retrofit was required to eliminate hands-on maintenance and difficulty in nozzle removal because of nozzle plugging. The retrofit design for this project demonstrates the solution of problems associated with thermal phenomena, structural supports, seismic requirements, remote handling and installations into extremely restricted spaces

  2. Proposed Atomic Energy of Canada Ltd. 99Mo waste calcination process

    International Nuclear Information System (INIS)

    Ramey, D.W.; Haas, P.A.; Malkemus, D.W.; McGinnis, C.P.; Meyers, E.S.; Patton, B.D.; Birdwell, J.F.; Jubin, R.T.; Coltharp, K.A.

    1994-10-01

    Atomic Energy of Canada Limited (AECL), at its Chalk River Laboratory, generates from 3000 to 5000 L/year of high-level fissile waste solution from the production of 99 Mo. In this Mo process, highly enriched uranium (93 wt % 235 U, total uranium basis) contained in uranium-aluminum alloy target rods is irradiated to produce the 99 Mo product. The targets are removed from the reactor and dissolved in a mercury nitrate-catalyzed reaction with nitric acid. The 99 Mo product is then recovered by passing the solution through an alumina (Al 2 O 3 ) column. During discussions with personnel from the Oak Ridge National Laboratory (ORNL) on September 10, 1992, the ORNL-developed technology formerly applied to the solidification of aqueous uranium waste (Consolidated Edison Uranium Solidification Program or CEUSP) was judged potentially applicable to the AECL 99 Mo waste. Under a Work-for-Others contract (no. ERD-92-1132), which began May 24, 1993, ORNL was tasked to determine the feasibility of applying the CEUSP (or a similar) calcination process to solidify AECL's 99 Mo waste for > 30 years of safe dry storage. This study was to provide sufficient detailed information on the applicability of a CEUSP-type waste solidification process to allow AECL to select the process which best suited its needs. As with the CEUSP process, evaporation of the waste and a simultaneously partial destruction of acid by reaction with formaldehyde followed by in situ waste can thermal denitration waste was chosen as the best means of solidification. Unlike the CEUSP material, the 99 Mo waste has a considerable number of problem volatile and semivolatile constituents which must be recovered in the off-gas treatment system. Mercury removal before calcination was seen as the best option

  3. Post treatment of high-level nuclear fuel wastes

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  4. Handbook of high-level radioactive waste transportation

    International Nuclear Information System (INIS)

    Sattler, L.R.

    1992-10-01

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

  5. The immobilization of High Level Waste Into Glass

    International Nuclear Information System (INIS)

    Aisyah; Martono, H.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    1985-03-01

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

  7. Techniques for the solidification of high-level wastes

    International Nuclear Information System (INIS)

    1977-01-01

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

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

    International Nuclear Information System (INIS)

    1982-03-01

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

  9. Final report on cermet high-level waste forms

    International Nuclear Information System (INIS)

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

    1981-08-01

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

  10. Evaluation of solidified high-level waste forms

    International Nuclear Information System (INIS)

    1981-01-01

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

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

    International Nuclear Information System (INIS)

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1997-09-01

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

  13. High-Level Waste System Process Interface Description

    International Nuclear Information System (INIS)

    D'Entremont, P.D.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    1987-11-01

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

  15. Radiation transport in high-level waste form

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    Kudo, Koji; Takada, Susumu; Kawanishi, Motoi

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    1976-06-01

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

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

    International Nuclear Information System (INIS)

    Lu Yingfa; Wu Yanchun; Luo Xianqi; Cui Yujun

    2006-01-01

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

  19. Cermet high level waste forms: a pregress report

    International Nuclear Information System (INIS)

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

    1978-06-01

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

  20. Environmental evaluation of alternatives for long-term management of Defense high-level radioactive wastes at the Idaho Chemical Processing Plant

    Energy Technology Data Exchange (ETDEWEB)

    1982-09-01

    The U.S. Department of Energy (DOE) is considering the selection of a strategy for the long-term management of the defense high-level wastes at the Idaho Chemical Processing Plant (ICPP). This report describes the environmental impacts of alternative strategies. These alternative strategies include leaving the calcine in its present form at the Idaho National Engineering Laboratory (INEL), or retrieving and modifying the calcine to a more durable waste form and disposing of it either at the INEL or in an offsite repository. This report addresses only the alternatives for a program to manage the high-level waste generated at the ICPP. 24 figures, 60 tables.

  1. Environmental evaluation of alternatives for long-term management of Defense high-level radioactive wastes at the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    1982-09-01

    The U.S. Department of Energy (DOE) is considering the selection of a strategy for the long-term management of the defense high-level wastes at the Idaho Chemical Processing Plant (ICPP). This report describes the environmental impacts of alternative strategies. These alternative strategies include leaving the calcine in its present form at the Idaho National Engineering Laboratory (INEL), or retrieving and modifying the calcine to a more durable waste form and disposing of it either at the INEL or in an offsite repository. This report addresses only the alternatives for a program to manage the high-level waste generated at the ICPP. 24 figures, 60 tables

  2. 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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

    1985-03-01

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

  5. Solidification of Savannah River Plant high-level waste

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  6. In vitro studies evaluating leaching of mercury from mine waste calcine using simulated human body fluids

    Science.gov (United States)

    Gray, John E.; Plumlee, Geoffrey S.; Morman, Suzette A.; Higueras, Pablo L.; Crock, James G.; Lowers, Heather A.; Witten, Mark L.

    2010-01-01

    In vitro bioaccessibility (IVBA) studies were carried out on samples of mercury (Hg) mine-waste calcine (roasted Hg ore) by leaching with simulated human body fluids. The objective was to estimate potential human exposure to Hg due to inhalation of airborne calcine particulates and hand-to-mouth ingestion of Hg-bearing calcines. Mine waste calcines collected from Hg mines at Almadén, Spain, and Terlingua, Texas, contain Hg sulfide, elemental Hg, and soluble Hg compounds, which constitute primary ore or compounds formed during Hg retorting. Elevated leachate Hg concentrations were found during calcine leaching using a simulated gastric fluid (as much as 6200 μg of Hg leached/g sample). Elevated Hg concentrations were also found in calcine leachates using a simulated lung fluid (as much as 9200 μg of Hg leached/g), serum-based fluid (as much as 1600 μg of Hg leached/g), and water of pH 5 (as much as 880 μg of Hg leached/g). The leaching capacity of Hg is controlled by calcine mineralogy; thus, calcines containing soluble Hg compounds contain higher leachate Hg concentrations. Results indicate that ingestion or inhalation of Hg mine-waste calcine may lead to increased Hg concentrations in the human body, especially through the ingestion pathway.

  7. Spent fuel and high-level radioactive waste storage

    International Nuclear Information System (INIS)

    Trigerman, S.

    1988-06-01

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

  8. Production and properties of solidified high-level waste

    International Nuclear Information System (INIS)

    Brodersen, K.

    1980-08-01

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

  9. Evaluation of conditioned high-level waste forms

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  10. Overview of high-level waste management accomplishments

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  11. Defense High Level Waste Disposal Container System Description Document

    International Nuclear Information System (INIS)

    Pettit, N. E.

    2001-01-01

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

  12. High level radioactive waste management facility design criteria

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    STAEHR, T.W.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    Baskin, K.P.

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    Cornman, W.R.

    1981-12-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  17. Development of high-level waste solidification technology 1

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-02-01

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

  18. Solidification of Savannah River Plant high level waste

    International Nuclear Information System (INIS)

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

    1981-11-01

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

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

    International Nuclear Information System (INIS)

    Gelin, R.

    1984-05-01

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

  20. Materials Science of High-Level Nuclear Waste Immobilization

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    Deokattey, Sangeeta; Bhanumurthy, K.

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  3. Disposal of high level and intermediate level radioactive wastes

    International Nuclear Information System (INIS)

    Flowers, R.H.

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    1992-01-01

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

  5. Site suitability criteria for solidified high level waste repositories

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  6. High-level radioactive waste repositories site selection plan

    International Nuclear Information System (INIS)

    Castanon, A.; Recreo, F.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Saire, D.E.

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-11-01

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

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

    International Nuclear Information System (INIS)

    Bernadzikowski, T.A.

    1981-01-01

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

  10. Spanish high level radioactive waste management system issues

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  11. Spanish high level radioactive waste management system issues

    International Nuclear Information System (INIS)

    Ulibarri, A.; Veganzones, A.

    1993-01-01

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

  12. High-level nuclear waste disposal: Ethical considerations

    International Nuclear Information System (INIS)

    Maxey, M.N.

    1985-01-01

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

  13. Retrieval System for Calcined Waste for the Idaho Cleanup Project - 12104

    Energy Technology Data Exchange (ETDEWEB)

    Eastman, Randy L.; Johnston, Beau A.; Lower, Danielle E. [CH2M-WG Idaho, LLC. The Idaho Cleanup Project at the Idaho National Laboratory (United States)

    2012-07-01

    This paper describes the conceptual approach to retrieve radioactive calcine waste, hereafter called calcine, from stainless steel storage bins contained within concrete vaults. The retrieval system will allow evacuation of the granular solids (calcine) from the storage bins through the use of stationary vacuum nozzles. The nozzles will use air jets for calcine fluidization and will be able to rotate and direct the fluidization or displacement of the calcine within the bin. Each bin will have a single retrieval system installed prior to operation to prevent worker exposure to the high radiation fields. The addition of an articulated camera arm will allow for operations monitoring and will be equipped with contingency tools to aid in calcine removal. Possible challenges (calcine bridging and rat-holing) associated with calcine retrieval and transport, including potential solutions for bin pressurization, calcine fluidization and waste confinement, are also addressed. The Calcine Disposition Project has the responsibility to retrieve, treat, and package HLW calcine. The calcine retrieval system has been designed to incorporate the functions and technical characteristics as established by the retrieval system functional analysis. By adequately implementing the highest ranking technical characteristics into the design of the retrieval system, the system will be able to satisfy the functional requirements. The retrieval system conceptual design provides the means for removing bulk calcine from the bins of the CSSF vaults. Top-down vacuum retrieval coupled with an articulating camera arm will allow for a robust, contained process capable of evacuating bulk calcine from bins and transporting it to the processing facility. The system is designed to fluidize, vacuum, transport and direct the calcine from its current location to the CSSF roof-top transport lines. An articulating camera arm, deployed through an adjacent access riser, will work in conjunction with the

  14. Canadian high-level radioactive waste management system issues

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  16. Vitrification of high-level radioactive and hazardous wastes

    International Nuclear Information System (INIS)

    Lutze, W.

    1993-12-01

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

  17. High-level waste management technology program plan

    International Nuclear Information System (INIS)

    Harmon, H.D.

    1995-01-01

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

  18. High-level waste management technology program plan

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, H.D.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Cotton, T.

    1985-01-01

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

  20. Defense High Level Waste Disposal Container System Description Document

    International Nuclear Information System (INIS)

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    Kaushik, C.P.

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    JANTZEN, CAROL

    2004-01-01

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

  3. Safe immobilization of high-level nuclear reactor wastes

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  4. High-level waste description, inventory and hazard

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-04-01

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

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

    International Nuclear Information System (INIS)

    Beale, H.

    1979-11-01

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

  7. High-level waste melter alternatives assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Calmus, R.B.

    1995-02-01

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

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

    International Nuclear Information System (INIS)

    Fawcett, R.

    1993-11-01

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

  9. High-level waste melter alternatives assessment report

    International Nuclear Information System (INIS)

    Calmus, R.B.

    1995-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-11-01

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

  11. On risk assessment of high level radioactive waste disposal

    International Nuclear Information System (INIS)

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

    1976-01-01

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

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

    International Nuclear Information System (INIS)

    1992-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1989-11-01

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

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

    International Nuclear Information System (INIS)

    1989-11-01

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

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

    International Nuclear Information System (INIS)

    1990-11-01

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

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

    International Nuclear Information System (INIS)

    Deffenbaugh, M.L.

    1997-01-01

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

  17. 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

  18. Safe disposal of high-level radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-10-01

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

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

    International Nuclear Information System (INIS)

    Ding Xiaobin; Wang Changhong; Zhu Hehua; Li Xiaojun

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  1. Storage of High Level Nuclear Waste in Germany

    Directory of Open Access Journals (Sweden)

    Dietmar P. F. Möller

    2007-01-01

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

  2. Safety of geologic disposal of high level radioactive waste

    International Nuclear Information System (INIS)

    Zaitsu, Tomohisa; Ishiguro, Katsuhiko; Masuda, Sumio

    1992-01-01

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

  3. Analysis of capital and operating costs associated with high level waste solidification processes

    International Nuclear Information System (INIS)

    Heckman, R.A.; Kniazewycz, B.G.

    1978-03-01

    An analysis was performed to evaluate the sensitivity of annual operating costs and capital costs of waste solidification processes to various parameters defined by the requirements of a proposed Federal waste repository. Five process methods and waste forms examined were: salt cake, spray calcine, fluidized bed calcine, borosilicate glass, and supercalcine multibarrier. Differential cost estimates of the annual operating and maintenance costs and the capital costs for the five HLW solidification alternates were developed

  4. Leaching behavior of simulated high-level waste glass

    International Nuclear Information System (INIS)

    Kamizono, Hiroshi

    1987-03-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Peterson, P.F.

    1999-01-01

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

  7. Status of the French nuclear high level waste disposal

    International Nuclear Information System (INIS)

    Sombret, C.

    1985-09-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

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

    International Nuclear Information System (INIS)

    Jing Xu

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

    The TRANSCOM (transportation tracking and communication) system is the US Department of Energy's (DOE's) real-time system for tracking shipments of spent fuel, high-level wastes, and other high-visibility shipments of radioactive material. The TRANSCOM system has been operational since 1988. The system was used during FY 1993 to track almost 100 shipments within the US DOE complex, and it is accessed weekly by 10 to 20 users

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

    International Nuclear Information System (INIS)

    Nash, C.

    1998-05-01

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

  13. Borosilicate glass as a matrix for immobilization of SRP high-level waste

    International Nuclear Information System (INIS)

    Wicks, G.G.

    1980-01-01

    Approximately 22 million gallons of high-level radioactive defense waste are currently being stored in large underground tanks located on the Savannah River Plant (SRP) site in Aiken, South Carolina. One option now being considered for long-term management of this waste involves removing the waste from the tanks, chemically processing the waste, and immobilizing the potentially harmful radionuclides in the waste into a borosilicate glass matrix. The technology for producing waste glass forms is well developed and has been demonstrated on various scales using simulated as well as radioactive SRP waste. Recently, full-scale prototypical equipment has been made operational at SRP. This includes both a joule-heated ceramic melter and an in-can melter. These melters are a part of an integrated vitrification system which is under evaluation and includes a spray calciner, direct liquid feed apparatus, and various elements of an off-gas system. Two of the most important properties of the waste glass are mechanical integrity and leachability. Programs are in progress at SRL aimed at minimizing thermally induced cracking by carefully controlling cooling cycles and using ceramic liners or coatings. The leachability of SRP waste glass has been studied under many different conditions and consistently found to be low. For example, the leachability of actual SRP waste glass was found to be 10 -6 to 10 -5 g/(cm 2 )(day) initially and decreasing to 10 -9 to 10 -8 g/(cm 2 )(day) after 100 days. Waste glass is also being studied under anticipated storage conditions. In brine at 90 0 C, the leachability is about 5 x 10 -8 g/(cm 2 )(day) after 60 days. The effects of other geological media including granite, basalt, shale, and tuff are also being studied as part of the multibarrier isolation system

  14. Strategic lessons in high-level waste management planning

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, Neil

    1999-07-01

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

  15. Strategic lessons in high-level waste management planning

    International Nuclear Information System (INIS)

    Chapman, Neil

    1999-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hustrulid, W

    1982-01-01

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

  17. ICPP radioactive liquid and calcine waste technologies evaluation final report and recommendation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

    Using a formalized Systems Engineering approach, the Latched Idaho Technologies Company developed and evaluated numerous alternatives for treating, immobilizing, and disposing of radioactive liquid and calcine wastes at the Idaho Chemical Processing Plant. Based on technical analysis data as of March, 1995, it is recommended that the Department of Energy consider a phased processing approach -- utilizing Radionuclide Partitioning for radioactive liquid and calcine waste treatment, FUETAP Grout for low-activity waste immobilization, and Glass (Vitrification) for high-activity waste immobilization -- as the preferred treatment and immobilization alternative.

  18. High level radioactive wastes: Considerations on final disposal

    International Nuclear Information System (INIS)

    Ciallella, Norberto R.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  20. Separation processes for high-level radioactive waste treatment

    International Nuclear Information System (INIS)

    Sutherland, D.G.

    1992-11-01

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

  1. A comparison of high-level waste form characteristics

    International Nuclear Information System (INIS)

    Salmon, R.; Notz, K.J.

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    1975-01-01

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

  3. Coupled processes in NRC high-level waste research

    International Nuclear Information System (INIS)

    Costanzi, F.A.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    1979-06-01

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

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

    International Nuclear Information System (INIS)

    Valentine, J.H.; Malone, K.; Schaus, P.S.

    1998-03-01

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

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

    International Nuclear Information System (INIS)

    Meyers, S.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  8. Managing commercial high-level radioactive waste: summary

    International Nuclear Information System (INIS)

    1982-04-01

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

  9. Engineering materials for high level radioactive waste repository

    International Nuclear Information System (INIS)

    Wen Zhijian

    2009-01-01

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

  10. Space augmentation of military high-level waste disposal

    International Nuclear Information System (INIS)

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

    1979-01-01

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

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

    International Nuclear Information System (INIS)

    Smith, C.F.

    1975-01-01

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

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

    International Nuclear Information System (INIS)

    Collins, K.J.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Kiser, D.L.

    1981-01-01

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

  14. Control of high level radioactive waste-glass melters

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    Zapp, P.E.

    1994-01-01

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

  16. Calcined solids storage facility closure study

    International Nuclear Information System (INIS)

    Dahlmeir, M.M.; Tuott, L.C.; Spaulding, B.C.

    1998-02-01

    The disposal of radioactive wastes now stored at the Idaho National Engineering and Environmental Laboratory is currently mandated under a open-quotes Settlement Agreementclose quotes (or open-quotes Batt Agreementclose quotes) between the Department of Energy and the State of Idaho. Under this agreement, all high-level waste must be treated as necessary to meet the disposal criteria and disposed of or made road ready to ship from the INEEL by 2035. In order to comply with this agreement, all calcined waste produced in the New Waste Calcining Facility and stored in the Calcined Solids Facility must be treated and disposed of by 2035. Several treatment options for the calcined waste have been studied in support of the High-Level Waste Environmental Impact Statement. Two treatment methods studied, referred to as the TRU Waste Separations Options, involve the separation of the high-level waste (calcine) into TRU waste and low-level waste (Class A or Class C). Following treatment, the TRU waste would be sent to the Waste Isolation Pilot Plant (WIPP) for final storage. It has been proposed that the low-level waste be disposed of in the Tank Farm Facility and/or the Calcined Solids Storage Facility following Resource Conservation and Recovery Act closure. In order to use the seven Bin Sets making up the Calcined Solids Storage Facility as a low-level waste landfill, the facility must first be closed to Resource Conservation and Recovery Act (RCRA) standards. This study identifies and discusses two basic methods available to close the Calcined Solids Storage Facility under the RCRA - Risk-Based Clean Closure and Closure to Landfill Standards. In addition to the closure methods, the regulatory requirements and issues associated with turning the Calcined Solids Storage Facility into an NRC low-level waste landfill or filling the bin voids with clean grout are discussed

  17. Calcined solids storage facility closure study

    Energy Technology Data Exchange (ETDEWEB)

    Dahlmeir, M.M.; Tuott, L.C.; Spaulding, B.C. [and others

    1998-02-01

    The disposal of radioactive wastes now stored at the Idaho National Engineering and Environmental Laboratory is currently mandated under a {open_quotes}Settlement Agreement{close_quotes} (or {open_quotes}Batt Agreement{close_quotes}) between the Department of Energy and the State of Idaho. Under this agreement, all high-level waste must be treated as necessary to meet the disposal criteria and disposed of or made road ready to ship from the INEEL by 2035. In order to comply with this agreement, all calcined waste produced in the New Waste Calcining Facility and stored in the Calcined Solids Facility must be treated and disposed of by 2035. Several treatment options for the calcined waste have been studied in support of the High-Level Waste Environmental Impact Statement. Two treatment methods studied, referred to as the TRU Waste Separations Options, involve the separation of the high-level waste (calcine) into TRU waste and low-level waste (Class A or Class C). Following treatment, the TRU waste would be sent to the Waste Isolation Pilot Plant (WIPP) for final storage. It has been proposed that the low-level waste be disposed of in the Tank Farm Facility and/or the Calcined Solids Storage Facility following Resource Conservation and Recovery Act closure. In order to use the seven Bin Sets making up the Calcined Solids Storage Facility as a low-level waste landfill, the facility must first be closed to Resource Conservation and Recovery Act (RCRA) standards. This study identifies and discusses two basic methods available to close the Calcined Solids Storage Facility under the RCRA - Risk-Based Clean Closure and Closure to Landfill Standards. In addition to the closure methods, the regulatory requirements and issues associated with turning the Calcined Solids Storage Facility into an NRC low-level waste landfill or filling the bin voids with clean grout are discussed.

  18. The high level and long lived radioactive wastes

    International Nuclear Information System (INIS)

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Blum, J.E.

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

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

  2. Pilot plant experience on high-level waste solidification and design of the engineering prototype VERA

    Energy Technology Data Exchange (ETDEWEB)

    Guber, W; Diefenbacher, W; Hild, W; Krause, H; Schneider, E; Schubert, G

    1972-11-01

    In the present paper the solidification process for highly active waste solutions as developed in the Karlsruhe Nuclear Research Center is presented. Its principal steps are: denitration, calcination in a spray calciner operated with superheated steam, melting of the calcine with appropriate additives to borosilicate glass in an induction-heated melting furnace. The operational experiences gained so far in the inactive 1:1 pilot plant are reported. Furthermore, a description is given of the projected multi-purpose experimental facility VERA 2 which is provided for processing the highly active waste solutions from the first German reprocessing plant WAK.

  3. Midwestern High-Level Radioactive Waste Transportation Project

    International Nuclear Information System (INIS)

    Dantoin, T.S.

    1990-12-01

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

  4. High level radioactive waste vitrification process equipment component testing

    International Nuclear Information System (INIS)

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

    1985-04-01

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

  5. 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

  6. Monitoring of geological repositories for high level radioactive waste

    International Nuclear Information System (INIS)

    2001-04-01

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

  7. High-level waste tank farm set point document

    International Nuclear Information System (INIS)

    Anthony, J.A. III.

    1995-01-01

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

  8. High-level waste tank farm set point document

    Energy Technology Data Exchange (ETDEWEB)

    Anthony, J.A. III

    1995-01-15

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

  9. National high-level waste systems analysis plan

    International Nuclear Information System (INIS)

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

    1995-05-01

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

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

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

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

  11. Thermo-aeraulics of high level waste storage facilities

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  12. Thermo-aeraulics of high level waste storage facilities

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  14. High-level radioactive waste isolation by incorporation in silicate rock

    International Nuclear Information System (INIS)

    Schwartz, L.L.; Cohen, J.J.; Lewis, A.E.; Braun, R.L.

    1978-01-01

    A number of technical possibilities for isolating high-level radioactive materials have been theoretically investigated at various times and places. Isolating such wastes deep underground to ensure long term removal from the biosphere is one such possibility. The present concept involves as a first step creating the necessary void space at considerable depth, say 2 to 5 km, in a very-low-permeability silicate medium such as shale. Waste in dry, calcined or vitrified form is then lowered into the void space, and the access hole or shaft sealed. Energy released by the radioactive decay raises the temperature to a point where the surrounding rock begins to melt. The waste is then dissolved in it. The extent of this melt region grows until the heat generated is balanced by conduction away from the molten zone. Resolidification then begins, and ends when the radioactive decay has progressed to the point that the temperature falls below the melting point of the rock-waste solution. Calculations are presented showing the growth and resolidification process. A nuclear explosion is one way of creating the void space. (author)

  15. Low-risk alternative waste forms for problematic high-level and long-lived nuclear wastes

    International Nuclear Information System (INIS)

    Stewart, M.W.A.; Begg, B.D.; Moricca, S.; Day, R.A.

    2006-01-01

    Full text: The highest cost component the nuclear waste clean up challenge centres on high-level waste (HLW) and consequently the greatest opportunity for cost and schedule savings lies with optimising the approach to HLW cleanup. The waste form is the key component of the immobilisation process. To achieve maximum cost savings and optimum performance the selection of the waste form should be driven by the characteristics of the specific nuclear waste to be immobilised, rather than adopting a single baseline approach. This is particularly true for problematic nuclear wastes that are often not amenable to a single baseline approach. The use of tailored, high-performance, alternative waste forms that include ceramics and glass-ceramics, coupled with mature process technologies offer significant performance improvements and efficiency savings for a nuclear waste cleanup program. It is the waste form that determines how well the waste is locked up (chemical durability), and the number of repository disposal canisters required (waste loading efficiency). The use of alternative waste forms for problematic wastes also lowers the overall risk by providing high performance HLW treatment alternatives. The benefits tailored alternative waste forms bring to the HLW cleanup program will be briefly reviewed with reference to work carried out on the following: The HLW calcines at the Idaho National Laboratory; SYNROC ANSTO has developed a process utilising a glass-ceramic combined with mature hot-isostatic pressing (HIP) technology and has demonstrated this at a waste loading of 80 % and at a 30 kg HIP scale. The use of this technology has recently been estimated to result in a 70 % reduction in waste canisters, compared to the baseline borosilicate glass technology; Actinide-rich waste streams, particularly the work being done by SYNROC ANSTO with Nexia Solutions on the Plutonium-residues wastes at Sellafield in the UK, which if implemented is forecast to result in substantial

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

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

    International Nuclear Information System (INIS)

    Mc Menamin, T.

    1989-01-01

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

  18. Midwestern High-Level Radioactive Waste Transportation Project

    International Nuclear Information System (INIS)

    1993-01-01

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

  19. Risk communication system for high level radioactive waste disposal

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  1. The principal radionuclides in high level radioactive waste management

    International Nuclear Information System (INIS)

    Mulyanto

    1998-01-01

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

  2. Calcination of Fluorinel-sodium waste blends using sugar as a feed additive (formerly WINCO-11879)

    International Nuclear Information System (INIS)

    Newby, B.J.; Thomson, T.D.; O'Brien, B.H.

    1992-06-01

    Methods were studied for using sugar as a feed additive for converting the sodium-bearing wastes stored at the Idaho Chemical Processing Plant into granular, free flowing solids by fluidized-bed calcination at 500 degrees C. All methods studied blended sodium-bearing wastes with Fluorinel wastes but differed in the types of sugar (sucrose or dextrose) that were added to the blend. The most promising sugar additive was determined to be sucrose, since it is converted more completely to inorganic carbon than is dextrose. The effect of the feed aluminum-to-alkali metal mole ratio on calcination of these blends with sugar was also investigated. Increasing the aluminum-to-alkali metal ratio from 0.6 to 1.0 decreased the calcine product-to-fines ratio from 3.0 to 1.0 and the attrition index from 80 to 15%. Further increasing the ratio to 1.25 had no effect

  3. Mechanism of ruthenium dioxide crystallization during high level waste vitrification

    International Nuclear Information System (INIS)

    Boucetta, H.

    2012-01-01

    Ruthenium, arising from the reprocessing of spent uranium oxide fuel, has a low solubility in glass melt. It crystallizes in the form of particles of RuO 2 of acicular or polyhedral morphology dispersed in fission product and actinides waste containment glass. Since the morphology of these particles strongly influences the physico-chemical properties, the knowledge and the control of their mechanism of formation are of major importance. The goal of this work is to determine the chemical reactions responsible for the formation of RuO 2 particles of acicular or polyhedral shape during glass synthesis. Using a simplification approach, the reactions between RuO 2 -NaNO 3 , and more complex calcine RuO 2 -Al 2 O 3 -Na 2 O and a sodium borosilicate glass are studied. In situ scanning electron microscopy and XANES at increasing temperatures are used to follow changes in composition, speciation and morphology of the ruthenium intermediate species. Those compounds are thoroughly characterised by SEM, XRD, HRTEM, and ruthenium K-edge X-ray absorption spectroscopy. This combined approach allows us to show that the ruthenium speciation modification during vitrification is the key of control of the morphology of RuO 2 particles in the glass. In particular, the formation of a specific intermediate compound (Na 3 RuO 4 ) is one of the main steps that lead to the precipitation of needle-shaped RuO 2 particles in the melt. The formation of polyhedral particles, on the contrary, results from the direct incorporation of RuO 2 crystals in the melt followed by an Ostwald ripening mechanism. (author) [fr

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

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

    International Nuclear Information System (INIS)

    Mendel, J.E.

    1984-08-01

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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  9. ATW system impact on high-level waste

    International Nuclear Information System (INIS)

    Arthur, E.D.

    1992-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mendel, J.E. (compiler)

    1984-08-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  12. High-level Waste Long-term management technology development

    International Nuclear Information System (INIS)

    Choi, Jong Won; Kang, C. H.; Ko, Y. K.

    2012-02-01

    The purpose of this project is to develop a long-term management system(A-KRS) which deals with spent fuels from domestic nuclear power stations, HLW from advanced fuel cycle and other wastes that are not admitted to LILW disposal site. Also, this project demonstrate the feasibility and reliability of the key technologies applied in the A-KRS by evaluating them under in-situ condition such as underground research laboratory and provide important information to establish the safety assessment and long-term management plan. To develop the technologies for the high level radioactive wastes disposal, demonstrate their reliability under in-situ condition and establish safety assessment of disposal system, The major objects of this project are the following: Ο An advanced disposal system including waste containers for HLW from advanced fuel cycle and pyroprocess has been developed. Ο Quantitative assessment tools for long-term safety and performance assessment of a radwaste disposal system has been developed. Ο Hydrological and geochemical investigation and interpretation methods has been developed to evaluate deep geological environments. Ο The THMC characteristics of the engineered barrier system and near-field has been evaluated by in-situ experiments. Ο The migration and retardation of radionuclides and colloid materials in a deep geological environment has been investigated. The results from this project will provide important information to show HLW disposal plan safe and reliable. The knowledge from this project can also contribute to environmental conservation by applying them to the field of oil and gas industries to store their wastes safe

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-01-10

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

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

    International Nuclear Information System (INIS)

    Weber, A.H.

    1999-01-01

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

  15. Thermal characteristics of rocks for high-level waste repository

    International Nuclear Information System (INIS)

    Shimooka, Kenji; Ishizaki, Kanjiro; Okamoto, Masamichi; Kumata, Masahiro; Araki, Kunio; Amano, Hiroshi

    1980-12-01

    Heat released by the radioactive decay of high-level waste in an underground repository causes a long term thermal disturbance in the surrounding rock mass. Several rocks constituting geological formations in Japan were gathered and specific heat, thermal conductivity, thermal expansion coefficient and compressive strength were measured. Thermal analysis and chemical analysis were also carried out. It was found that volcanic rocks, i.e. Andesite and Basalt had the most favorable thermal characteristics up to around 1000 0 C and plutonic rock, i.e. Granite had also favorable characteristics under 573 0 C, transition temperature of quartz. Other igneous rocks, i.e. Rhyolite and Propylite had a problem of decomposition at around 500 0 C. Sedimentary rocks, i.e. Zeolite, Tuff, Sandstone and Diatomite were less favorable because of their decomposition, low thermal conductivity and large thermal expansion coefficient. (author)

  16. Transmutation of high-level radioactive waste - Perspectives

    CERN Document Server

    Junghans, Arnd; Grosse, Eckart; Hannaske, Roland; Kögler, Toni; Massarczyk, Ralf; Schwengner, Ronald; Wagner, Andreas

    2014-01-01

    In a fast neutron spectrum essentially all long-lived actinides (e.g. Plutonium) undergo fission and thus can be transmuted into generally short lived fission products. Innovative nuclear reactor concepts e.g. accelerator driven systems (ADS) are currently in development that foresee a closed fuel cycle. The majority of the fissile nuclides (uranium, plutonium) shall be used for power generation and only fission products will be put into final disposal that needs to last for a historical time scale of only 1000 years. For the transmutation of high-level radioactive waste a lot of research and development is still required. One aspect is the precise knowledge of nuclear data for reactions with fast neutrons. Nuclear reactions relevant for transmutation are being investigated in the framework of the european project ERINDA. First results from the new neutron time-of-flight facility nELBE at Helmholtz-Zentrum Dresden-Rossendorf will be presented.

  17. High-level radioactive-waste-disposal investigations in Texas

    International Nuclear Information System (INIS)

    Smith, R.D.

    1983-01-01

    The Texas Energy and Natural Resources Advisory Council (TENRAC) was designated in 1980 to coordinate the interaction between the State of Texas and the federal government relating to the high-level radioactive waste disposal issue. This report was prepared to summarize the many aspects of that issue with particular emphasis on the activities in Texas. The report is intended to provide a comprehensive introduction for individuals with little or no previous exposure to the issue and to provide a broader perspective for those individuals who have addressed specific aspects of the issue but have not had the opportunity to study it in a broader context. Following the introduction, contents of this report are as follows: (1) general status of major repository siting investigations in the US; (2) detailed review of Texas studies; (3) possible facilities to be sited in Texas; (4) current Texas policy; (5) federal regulations; and (6) federal legislation. 9 figures, 2 tables

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

    International Nuclear Information System (INIS)

    Gannon, R.E.

    1993-01-01

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

  19. Geology of high-level nuclear waste disposal

    International Nuclear Information System (INIS)

    Roxburgh, I.S.

    1988-01-01

    The concept of geological disposal is set out by describing the major rock types in terms of their ability to isolate high-level nuclear waste. The advantages and problems posed by particular rock formations are explored and the design and construction of geological repositories is considered, along with the methods used to estimate their safety. It gives special consideration to the use of sea-covered rock and sediment as well as the on-land situation. Throughout the book the various principles and problems inherent in geological disposal are explained and illustrated by reference to a multitude of European and North American case studies, backed up by a large number of tables, figures and an extensive bibliography

  20. Risk assessment methodology for Hanford high-level waste tanks

    International Nuclear Information System (INIS)

    Bott, T.F.; Mac Farlane, D.R.; Stack, D.W.; Kindinger, J.

    1992-01-01

    A methodology is presented for applying Probabilistic Safety Assessment techniques to quantification of the health risks posed by the high-level waste (HLW) underground tanks at the Department of Energy's Hanford reservation. This methodology includes hazard screening development of a list of potential accident initiators, systems fault trees development and quantification, definition of source terms for various release categories, and estimation of health consequences from the releases. Both airborne and liquid pathway releases to the environment, arising from aerosol and spill/leak releases from the tanks, are included in the release categories. The proposed methodology is intended to be applied to a representative subset of the total of 177 tanks, thereby providing a baseline risk profile for the HLW tank farm that can be used for setting clean-up/remediation priorities. Some preliminary results are presented for Tank 101-SY

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

    International Nuclear Information System (INIS)

    WASHENFELDER DJ

    2008-01-01

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

  2. Underground excavation methods for a high-level waste repository

    International Nuclear Information System (INIS)

    Peshel, J.; Gupta, D.; Nataraja, M.

    1990-01-01

    This paper reports on rock excavation methods for a High-Level Waste repository that should be selected to limit the potential for creating preferential pathways for groundwater to travel to the waste packages or for radionuclides to migrate to the accessible environment. The use of water and other foreign substances should be controlled so that the repository performance is not compromised. The excavated openings should remain stable so that operations can be carried out safely and the retrievability option maintained. As per the current conceptual designs presented by the Department of Energy, the exploratory shaft facility becomes a part of the repository if the Yucca Mountain site is found suitable for repository development. Therefore, the methods of constructing the underground openings should be compatible with the performance requirements for the repository. Also, the degree of damage to the rock surrounding the openings and the extent of the damage zone should not preclude adequate site characterization. The ESf construction and operation should be compatible with the site data gathering activities, such as geological, thermomechanical, hydrological and geochemical testing

  3. The disposal of high-level radioactive waste. Vol. 1

    International Nuclear Information System (INIS)

    Parker, F.L.; Broshears, R.E.; Pasztor, J.

    1984-01-01

    The Beijer Institute received request from the Swedish Board for Spent Nuclear Fuel (Naemnden for Anvaent Kaernbraensle - NAK) to undertake an international review of the major programmes which were currently making arrangements for the future disposal of high-level radioactive wastes and spent nuclear fuel. The request was accepted, a detailed proposal was worked out and agreed to by NAK, for a critical technical review which concentrated on the following three main tasks: 1. a 'state-of-the-art' review of selected ongoing disposal programmes, both national and international; 2. an assessment of the scientific and technical controversies involved, and 3. recommendations for further research in this field. This review work was to be built on a survey of the available technical literature which was to serve as a basis for a series of detailed interviews, consultations and discussions with scientific and technical experts in Japan, Canada, USA, Belgium, Federal Republic of Germany, France, Switzerland and the United Kingdom. This first volume contains: disposal options; review of the state-of-the-art (international activities, national programs); analysis of waste disposal systems. (orig./HP)

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

    International Nuclear Information System (INIS)

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

    1996-03-01

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

  5. Safety of geological disposal of high-level waste

    International Nuclear Information System (INIS)

    Ohe, Toshiaki; Tsukamoto, Masaki

    1989-01-01

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

  6. The IFR pyroprocessing for high-level waste minimization

    International Nuclear Information System (INIS)

    Laidler, J.J.

    1993-01-01

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

  7. Control of high-level radioactive waste-glass melters

    International Nuclear Information System (INIS)

    Bickford, D.F.; Coleman, C.J.

    1990-01-01

    The Defense Waste Processing Facility (DWPF) will immobilize Savannah River Site High Level Waste as a durable borosilicate glass for permanent disposal in a repository. The DWPF will be controlled based on glass composition. The following discussion is a preliminary analysis of the capability of the laboratory methods that can be used to control the glass composition, and the relationships between glass durability and glass properties important to glass melting. The glass durability and processing properties will be controlled by controlling the chemical composition of the glass. The glass composition will be controlled by control of the melter feed transferred from the Slurry Mix Evaporator (SME) to the Melter Feed Tank (MFT). During cold runs, tests will be conducted to demonstrate the chemical equivalence of glass sampled from the pour stream and glass removed from cooled canisters. In similar tests, the compositions of glass produced from slurries sampled from the SME and MFT will be compared to final product glass to determine the statistical relationships between melter feed and glass product. The total error is the combination of those associated with homogeneity in the SME or MFT, sampling, preparation of samples for analysis, instrument calibration, analysis, and the composition/property model. This study investigated the sensitivity of estimation of property data to the combination of variations from sampling through analysis. In this or a similar manner, the need for routine glass product sampling will be minimized, and glass product characteristics will be assured before the melter feed is committed to the melter

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

    Energy Technology Data Exchange (ETDEWEB)

    J.A. Ziegler

    2000-11-20

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

  9. Immobilization of high-level wastes into sintered glass: 2

    International Nuclear Information System (INIS)

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

    1987-01-01

    High level radioactive wastes are immobilized into borosilicate glasses. Experiences with the variety VG 98/12 (SiO 2 , TiO 2 , Al 2 O 3 , B 2 O 3 , MgO, CaO, Na 2 O) are described. The pressing was performed in a matrix of 12.7 mm diameter, the walls of which were lubricated with sterotex dissolved in Cl 4 C. The sintering was made in an horizontal electric furnace in air atmosphere at temperatures between 500 and 600 deg C. It was observed that the maximum density occurs at 605 deg C. Comparing both the hot and the cold pressing process, it is concluded that: 1) In spite of requiring more complex equipment the hot pressing process has the advantage that lower pressures are applied, with the consequent obtainment of waste blocks with larger diameters, and 2) it is advisable that pressing process should be performed in the definitive can. (M.E.L.) [es

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    D.C. Richardson

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    D.C. Richardson

    2003-03-19

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

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

    Energy Technology Data Exchange (ETDEWEB)

    WILLIS, W.L.

    2000-06-15

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

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

    International Nuclear Information System (INIS)

    WILLIS, W.L.

    2000-01-01

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

  15. Final repositories for high-level radioactive waste; Endlagerung hochradioaktiver Abfaelle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2015-10-15

    The brochure on final repositories for high-level radioactive waste covers the following issues: What is the origin of radioactive wastes? How large are the waste amounts? What is going to happen with the wastes? What is the solution for the Waste disposal? A new site search is started. Safety requirements for the final disposal of high-level radioactive wastes. Comparison of host rocks. Who is responsible and who will pay? Final disposal of high-level radioactive wastes worldwide. Short summary: History of the search for a final repository for high-level radioactive wastes in Germany.

  16. 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

  17. Biodiesel production from waste cooking oil using calcined scallop shell as catalyst

    International Nuclear Information System (INIS)

    Sirisomboonchai, Suchada; Abuduwayiti, Maidinamu; Guan, Guoqing; Samart, Chanatip; Abliz, Shawket; Hao, Xiaogang; Kusakabe, Katsuki; Abudula, Abuliti

    2015-01-01

    Highlights: • Calcined scallop shell was used as low-cost and effective catalyst for biodiesel production. • BDF yield from waste cooking oil reached 86% at 65 °C with a catalyst loading amount of 5 wt%. • Calcined scallop shell showed good reusability. • Calcium glyceroxide played an important role on the reusability of calcined scallop shell. • Water in the waste cooking oil had negative effect on the catalytic activity of calcined scallop shell. - Abstract: Transesterification of waste cooking oil (WCO) and methanol by using calcined scallop shell (CSS) as catalyst was carried out in a closed system for biodiesel fuel (BDF) production. It is found that the optimum calcination temperature for the preparation of CSS was 1000 °C. The effects of transesterification temperature, reaction time, methanol/oil molar ratio and catalyst loading amount on the BDF yield were investigated. Compared with the commercial CaO, CSS showed higher catalytic activity and the BDF yield reached 86% at 65 °C with a catalyst loading amount of 5 wt% (WCO basis) and a reaction time of 2 h. The catalyst was reused for 5 cycles whilst the BDF yield decreased 23%. It is found that CaO in CSS was transferred to calcium glyceroxide after the transesterification reaction, and calcium glyceroxide also showed good catalytic activity and reusability. Furthermore, Water content in WCO had negative effect on BDF yield. It is found that BDF yield reduced 15% due to the occurring of saponification when the water content was increased from 0.64% to 2.48%. It is expected that CCS can be used as an alternative and cheap catalyst for the biodiesel production

  18. Pecularities of carrying out radioactive wastes vitrification process without preliminary calcination of wastes

    International Nuclear Information System (INIS)

    Konstantinovich, A.A.; Kulichenko, V.V.; Bel'tyukov, V.A.; Nikiforov, A.S.; Nikipelov, B.V.; Stepanov, S.E.; Baskov, L.I.; Kulakov, S.I.

    1978-01-01

    Vitrification technology is considered for liquid radioactive wastes by means of electric furnace where heating of glass-paste is done by electric current passing through the melt. Continious process of gehydration, calcination and vitrification is going on in one apparatus. Testing if the method has been performed by use of a model solution, containing sodium and aluminium nitrates. To obtain phosphoric acid has been added into the solution. Lay-out of the device and its description as well as technical parameters of the electric furnace are given. The results are stated for determination of the optimum operation conditions for the device. To reduce entrainment of solid components, molasses has been added in the solution. Parameters are given for the process of the solution containing 80 g/l molasses processing. It has been shown that edding molasses to the solution permitted to reduse power consumption of the process due to the heat generation during oxidation-reduction reaction on the melt surface. The results are given for investigations of the nitrogen oxides catching in scrubbers. These results have shown that introduction of molasses reduces nitrigen oxides concentration. The results of the experimental works have shown the possibility of the continious process of dehydration, calcination and vitrification in single device with application of remote control and monitoring by means of automatics. (I.T.) [ru

  19. Study on the development of safety regulations for geological disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    Wei Fangxin

    2012-01-01

    The development of regulations under Regulations on Safety Management of Radioactive Waste has become necessary as the issuance of it. The regulations related to geological disposal of high-level radioactive waste can promote the progress of research and development on geological disposal of high-level radioactive waste in China. This paper has present suggestions on development of regulations on geological disposal of high-level radioactive waste by analyzing development of safety regulations on geological disposal of high-level radioactive waste in foreign countries and problems occurred in China and discussed important issues related to the development of safety regulations on geological disposal of high-level radioactive waste. (author)

  20. Alternative calcination development status report

    International Nuclear Information System (INIS)

    Boardman, R.D.

    1997-12-01

    The Programmatic Spent Nuclear Fuel and (INEEL) Environmental Restoration and Waste Management Programs Environmental Impact Statement Record of Decision, dated June 1, 1995, specifies that high-level waste stored in the underground tanks at the ICPP continue to be calcined while other options to treat the waste are studied. Therefore, the High-Level Waste Program has funded a program to develop new flowsheets to increase the liquid waste processing rate. Simultaneously, a radionuclide separation process, as well as other options, are also being developed, which will be compared to the calcination treatment option. Two alternatives emerged as viable candidates; (1) elevated temperature calcination (also referred to as high temperature calcination), and (2) sugar-additive calcination. Both alternatives were determined to be viable through testing performed in a lab-scale calcination mockup. Subsequently, 10-cm Calciner Pilot Plant scoping tests were successfully completed for both flowsheets. The results were compared to the standard 500 C, high-ANN flow sheet (baseline flowsheet). The product and effluent streams were characterized to help elucidate the process chemistry and to investigate potential environmental permitting issues. Several supplementary tests were conducted to gain a better understanding of fine-particles generation, calcine hydration, scrub foaming, feed makeup procedures, sugar/organic elimination, and safety-related issues. Many of the experiments are only considered to be scoping tests, and follow-up experiments will be required to establish a more definitive understanding of the flowsheets. However, the combined results support the general conclusion that flowsheet improvements for the NWCF are technically viable

  1. High-level waste issues and resolutions document

    International Nuclear Information System (INIS)

    1994-05-01

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

  2. The precautionary principle and high-level nuclear waste policy

    International Nuclear Information System (INIS)

    Frishman, S.

    1999-01-01

    The 'Precautionary Principle' has grown from the broadening observation that there is compelling evidence that damage to humans and the world-wide environment is of such a magnitude and seriousness that new principles for conducting human activities are necessary. One of the various statements of the Precautionary Principle is: when an activity raises threats of harm to human health or the environment, precautionary measures should be taken even if some cause and effect relationships are not fully established scientifically. The use of a precautionary principle was a significant recommendation emerging from the 1992 United Nations Conference on Environment and Development, held in Rio de Janeiro, Brazil, and it is gaining acceptance in discussions ranging from global warming to activities that affect the marine environment, and far beyond. In the US high-level nuclear waste policy, there is a growing trend on the part of geologic repository proponents and regulators to shift the required safety evaluation from a deterministic analysis of natural and engineered barriers and their interactions to risk assessments and total system waste containment and isolation performance assessment. This is largely a result of the realisation that scientific 'proof' of safety cannot be demonstrated to the level repository proponents have led the American public to expect. Therefore, they are now developing other methods in an attempt to effectively lower the repository safety expectations of the public. Implicit in this shift in demonstration of 'proof' is that levels of uncertainty far larger than those generally taken as scientifically acceptable must be accepted in repository safety, simply because greater certainty is either too costly, in time and money, or impossible to achieve at the potential Yucca Mountain repository site. In the context of the Precautionary Principle, the repository proponent must bear the burden of providing 'Acceptable' proof, established by an open

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

    International Nuclear Information System (INIS)

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

    1994-01-01

    The disposition of actinide elements released from high-level waste glasses into a tuff groundwater in laboratory tests at 90 degrees C at various glass surface area/leachant volume ratios (S/V) between dissolved, suspended, and sorbed fractions has been measured. While the maximum release of actinides is controlled by the corrosion rate of the glass matrix, their solubility and sorption behavior affects the amounts present in potentially mobile phases. Actinide solubilities are affected by the solution pH and the presence of complexants released from the glass, such as sulfate, phosphate, and chloride, radiolytic products, such as nitrate and nitrite, and carbonate. Sorption onto inorganic colloids formed during lass corrosion may increase the amounts of actinides in solution, although subsequent sedimentation of these colloids under static conditions leads to a significant reduction in the amount of actinides in solution. The solution chemistry and observed actinide behavior depend on the S/V of the test. Tests at high S/V lead to higher pH values, greater complexant concentrations, and generate colloids more quickly than tests at low S/V. The S/V also affects the rate of glass corrosion

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  5. Site suitability criteria for solidified high level waste repositories

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  6. Concentration of High Level Radioactive Liquid Waste. Basic data acquisition

    Energy Technology Data Exchange (ETDEWEB)

    Juvenelle, A.; Masson, M.; Garrido, M.H. [DEN/VRH/DRCP/SCPS/LPCP, BP 17171 - 30207 Bagnols sur Ceze Cedex (France)

    2008-07-01

    Full text of publication follows: In order to enhance its knowledge about the concentration of high level liquid waste (HLLW) from the nuclear fuel reprocessing process, a program of studies was defined by Cea. In a large field of acidity, it proposes to characterize the concentrated solution and the obtained precipitates versus the concentration factor. Four steps are considered: quantification of the salting-out effect on the concentrate acidity, acquisition of solubility data, precipitates characterisation versus the concentration factor through aging tests and concentration experimentation starting from simulated fission products solutions. The first results, reported here, connect the acidity of the concentrated solution to the concentration factor and allow us to precise the field of acidity (4 to 12 N) for the next experiments. In this field, solubility data of various elements (Ba, Sr, Zr...) are separately measured at room temperature, in nitric acid in a first time, then in the presence of various species present in medium (TBP, PO{sub 4}{sup 3-}). The reactions between these various elements are then investigated (formation of insoluble mixed compounds) by following the concentration cations in solution and characterising the precipitates. (authors)

  7. Why consider subseabed disposal of high-level nuclear wastes

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  8. Chemical reactivity of precursor materials during synthesis of glasses used for conditioning high-level radioactive waste: Experiments and models

    International Nuclear Information System (INIS)

    Monteiro, A.

    2012-01-01

    The glass used to store high-level radioactive waste is produced by reaction of a solid waste residue and a glassy precursor (glass frit). The waste residue is first dried and calcined (to lose water and nitrogen respectively), then mixed with the glass frit to enable vitrification at high temperature. In order to obtain a good quality glass of constant composition upon cooling, the chemical reactions between the solid precursors must be complete while in the liquid state, to enable incorporation of the radioactive elements into the glassy matrix. The physical and chemical conditions during glass synthesis (e.g. temperature, relative proportions of frit and calcine, amount of radioactive charge) are typically empirically adjusted to obtain a satisfactory final product. The aim of this work is to provide new insights into the chemical and physical interactions that take place during vitrification and to provide data for a mathematical model that has been developed to simulate the chemical reactions. The consequences of the different chemical reactions that involve solid, liquid and gaseous phases are described (thermal effects, changes in crystal morphology and composition, variations in melt properties and structure). In a first series of experiments, a simplified analogue of the calcine (NaNO 3 -Al 2 O 3 ± MoO 3 /Nd 2 O 3 ) has been studied. In a second series of experiments, the simplified calcines have been reacted with a simplified glass frit (SiO 2 -Na 2 O-B 2 O 3 -Al 2 O 3 ) at high temperature. The results show that crystallization of the calcine may take place before interaction with the glass frit, but that the reactivity with the glass at high temperature is a function of the nature and stoichiometry of the crystalline phases which form at low temperature. The results also highlight how the mixing of the starting materials, the physical properties of the frit (viscosity, glass transition temperature) and the Na 2 O/Al 2 O 3 of the calcine but also its

  9. Redox calcination study of Synroc D powder containing simulated SRL waste

    International Nuclear Information System (INIS)

    Chen, C.

    1982-01-01

    According to Ringwood [A.E. Ringwood, W. Sinclair, and G.M. McLaughlin, Nuclear Waste Immobilization, Lawrence Livermore Laboratory, Livermore, Rept. UCRL-15147 (1979)], the iron oxidation state is important in controlling, the spinel mineralogy and composition if the amount of titania (TiO 2 ) consumed in spinel formation is to be minimized in favor of the formation of the Synroc phases, zirconolite, perovskite, and nepheline. In our redox calcination studies we observed that the iron oxidation state of FeO/Fe 2 O 3 can be controlled by the redoxcalcining atmosphere. In a CO atmosphere, the oxidation state was reduced to less than 7 wt % Fe 2 O 3 . With appropriate CO 2 /CO gas mixtures the resultant iron oxidation states were in the range of 45 to 59 wt % Fe 2 O 3 . Direct rotary redox calcination of spray dried powder at 600 0 C, without prior air calcination, showed increased redox efficiency when compared to powder that had been previously air calcined at 650 0 C. We believe this is caused by a reduction in particle size. Rotary calcination at 800 0 C in argon has no measurable reduction affect on the iron oxidation state of Synroc D powder

  10. Characterization and vitrification of Hanford radioactive high level wastes

    International Nuclear Information System (INIS)

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

    1991-01-01

    Radioactive Neutralized Current Acid Waste (NCAW) samples from the Hanford waste tanks have been chemically, radiochemically and physically characterized. The wastes were processed according to the Hanford Waste vitrification Plant (HWVP) flowsheet, and characterized after each process step. The waste glasses were sectioned and leach tested. Chemical, radiochemical and physical properties of the waste will be presented and compared to nonradioactive simulant data and the HWVP reference composition and properties

  11. Continuous Calcination of High-Level Radioactive Wastes by Means of a Rotary-Ball Kiln; Calcination Continue de Dechets de Haute Activite dans un Four Rotatif a Billes; 041d 0415 041f 0420 0415 0420 042b 0412 041d 041e 0415 041a 0410 041b 042c 0426 0418 041d 0418 0420 041e 0412 0410 041d 0418 0415 0412 042b 0421 041e 041a 041e 0410 041a 0422 0418 0412 041d 042b 0425 041e 0422 0425 041e 0414 041e 0412 0412 041e 0412 0420 0410 0429 0410 042e 0429 0415 0419 0421 042f 0428 0410 0420 041e 0412 041e 0419 041f 0415 0427 0418 ; Calcinacion Continua de Desechos de Elevada Radiactividad en un Horno de Bolas Rotatorio

    Energy Technology Data Exchange (ETDEWEB)

    Regan, W. H.; Hatch, L. P.; Domish, R. F. [Brookhaven National Laboratory, Upton, Long Island, NY (United States)

    1963-02-15

    The calcination of high-level, high salt content, aqueous wastes to convert the bulk salts to stable solid form offers important advantages in ultimate waste disposal, namely, major reduction in waste volume and general simplification of long-range waste-storage problems. Among the types of wastes commonly considered for calcination are those produced in the aqueous reprocessing of aluminium alloy and zirconium-alloy fuels and in the reprocessing of fuels clad with stainless steel or having stainless-steel matrices. Decomposition of the bulk salts thus involved, namely, aluminium nitrate, zirconium fluoride and ferric nitrate to their respective oxides introduces a number of serious operations problems many of which stem from the formation of highly radioactive solids with substantial proportions in the small, or low micron, range. One promising type of calciner, which is under development at the Brookhaven National Laboratory operates on the principle of a rotary-ball kiln. Basically, the unit consists of a slowly rotating horizontal tube or kiln externally heated by means of electric resistance heaters. Feed is introduced into the kiln by means of a distribution nozzle and falls on a hot, shallow bed of metal balls, which aid in heat transfer and prevent the formation of large agglomerates during calcination. The calcined solids overflow from the discharge end of the kiln into a powder receiver. The calciner is designed for continuous operation, to give good transfer of heat to the interior of the kiln and to reduce problems of controlling radioactive dust to a bare minimum. Since no non-condensable gases are added to the system, excellent dust control is accomplished by direct condensation of off-gases followed by passing the small non-condensable portion through a scrubber and absolute filter. An engineering scale calciner has been constructed at Brookhaven and is at present being used to carry out equipment and process development studies. The rotary kiln is 8

  12. Waste package designs for disposal of high-level waste in salt formations

    International Nuclear Information System (INIS)

    Basham, S.J. Jr.; Carr, J.A.

    1984-01-01

    In the United States of America the selected method for disposal of radioactive waste is mined repositories located in suitable geohydrological settings. Currently four types of host rocks are under consideration: tuff, basalt, crystalline rock and salt. Development of waste package designs for incorporation in mined salt repositories is discussed. The three pertinent high-level waste forms are: spent fuel, as disassembled and close-packed fuel pins in a mild steel canister; commercial high-level waste (CHLW), as borosilicate glass in stainless-steel canisters; defence high-level waste (DHLW), as borosilicate glass in stainless-steel canisters. The canisters are production and handling items only. They have no planned long-term isolation function. Each waste form requires a different approach in package design. However, the general geometry and the materials of the three designs are identical. The selected waste package design is an overpack of low carbon steel with a welded closure. This container surrounds the waste forms. Studies to better define brine quantity and composition, radiation effects on the salt and brines, long-term corrosion behaviour of the low carbon steel, and the leaching behaviour of the spent fuel and borosilicate glass waste forms are continuing. (author)

  13. Incorporation of high-level nuclear waste in gel spheres

    International Nuclear Information System (INIS)

    Robinson, S.M.; Arnold, W.D.; Bond, W.D.; Angelini, P.; Stinton, D.P.

    1981-01-01

    Waste sludge is incorporated in gel spheres by the method of internal gelation. Gel spheres containing up to 90 wt % waste have been produced from defense and commercial wastes. A generic cesium-bearing waste form has been developed. Pyrolytic carbon and SiC coatings reduce the leachability of all tested articles to the detection limits

  14. Actinides and fission products partitioning from high level liquid waste

    International Nuclear Information System (INIS)

    Yamaura, Mitiko

    1999-01-01

    The presence of small amount of mixed actinides and long-lived heat generators fission products as 137 Cs and 90 Sr are the major problems for safety handling and disposal of high level nuclear wastes. In this work, actinides and fission products partitioning process, as an alternative process for waste treatment is proposed. First of all, ammonium phosphotungstate (PWA), a selective inorganic exchanger for cesium separation was chosen and a new procedure for synthesizing PWA into the organic resin was developed. An strong anionic resin loaded with tungstate or phosphotungstate anion enables the precipitation of PWA directly in the resinous structure by adding the ammonium nitrate in acid medium (R-PWA). Parameters as W/P ratio, pH, reactants, temperature and aging were studied. The R-PWA obtained by using phosphotungstate solution prepared with W/P=9.6, 9 hours digestion time at 94-106 deg C and 4 to 5 months aging time showed the best capacity for cesium retention. On the other hand, Sr separation was performed by technique of extraction chromatography, using DH18C6 impregnated on XAD7 resin as stationary phase. Sr is selectively extracted from acid solution and >99% was recovered from loaded column using distilled water as eluent. Concerning to actinides separations, two extraction chromatographic columns were used. In the first one, TBP(XAD7) column, U and Pu were extracted and its separations were carried-out using HNO 3 and hydroxylamine nitrate + HNO 3 as eluent. In the second one, CMP0-TBP(XAD7) column, the actinides were retained on the column and the separations were done by using (NH 4 ) 2 C 2 O 4 , DTPA, HNO 3 and HCl as eluent. The behavior of some fission products were also verified in both columns. Based on the obtained data, actinides and fission products Cs and Sr partitioning process, using TBP(XAD7) and CMP0-TBP(XAD7) columns for actinides separation, R-PWA column for cesium retention and DH18C6(XAD7) column for Sr isolation was performed

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

    International Nuclear Information System (INIS)

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

    1996-10-01

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

  16. High-level radioactive waste disposal in the deep ocean

    International Nuclear Information System (INIS)

    Hill, H.W.

    1977-01-01

    A joint programme has begun between the Fisheries Laboratory, Lowestoft and the Institute of Oceanographic Sciences, Wormley to study the dispersion of radioactivity in the deep ocean arising from the possible dumping of high level waste on the sea bed in vitrified-glass form which would permit slow leakage over a long term scale. The programme consists firstly of the development of a simple diffusion/advection model for the dispersion of radioactivity in a closed and finite ocean, which overcomes many of the criticisms of the earlier model proposed by Webb and Morley. Preliminary results from this new model are comparable to those of the Webb-Morley model for radio isotopes with half-lives of 10-300 years but are considerably more restrictive outside this range, particularly for those which are much longer-lived. The second part of the programme, towards which the emphasis is directed, concerns the field programme planned to measure the advection and diffusion parameters in the deeper layers of the ocean to provide realistic input parameters to the model and increase our fundamental understanding of the environment in which the radioactive materials may be released. The first cruises of the programme will take place in late 1976 and involve deep current meter deployments and float dispersion experiments around the present NEA dump site with some sediment sampling, so that adsorption experiments can be started on typical deep sea sediments. The programme will expand the number of long-term deep moored stations over the next five years and include further float experiments, CTD profiling, and other physical oceanography. In the second half of the 5-year programme, attempts will be made to measure diffusion parameters in the deeper layers of the ocean using radioactive tracers

  17. Interaction of cementitious materials with high-level waste

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  18. Final conditioning of high-level liquid radioactive waste

    International Nuclear Information System (INIS)

    Krause, H.

    1981-01-01

    Problems of the solidification of HLLW from the reprocessing stage are discussed. The matrix for embedding the 37 different fission products and the 5 actinides is of great importance in this context. In addition to glass, there is a number of other candidate materials, as e.g. 'super calcine' or 'synroc', which however need further research and experimental studies, as results so far are not so satisfactory. Vitrification is the most advanced technical method currently, and is practically applied in France's AVM installation and further tested in simulation experiments. A variant developed by KfK, vitrification in a ceramic melter with electrode heating, is explained by this paper. (RB) [de

  19. Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, R.P. [ed.

    1995-03-01

    This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM.

  20. Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results

    International Nuclear Information System (INIS)

    Rechard, R.P.

    1995-03-01

    This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM

  1. French industrial plant AVM for continuous vitrification of high level radioactive wastes

    International Nuclear Information System (INIS)

    Bonniaud, Roger; Sombret, Claude; Barbe, Alain

    1975-01-01

    The A.V.M. plant is a continuous process plant now under construction at Marcoule and intended for vitrifying the whole of fission product solutions from the C.E.A. (Commissariat a l'Energie Atomique) - Marcoule reprocessing plant. The outset of the construction took place in the second 1974 half year; the first radioactive run is scheduled in July 1977. The two steps of the process are shown: first a continuous calcination then a continuous glass making from the calcined product and suitable additives. The plant consists in two parts: vitrification and storage facilities. Some wastes will be continuously produced day after day due to gas clean up and worn out materials. Characteristics of the solutions processed, calcined products, glass composition, and expected liquid wastes are given in tables [fr

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

    International Nuclear Information System (INIS)

    Jenkins, C.F.

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

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

  4. Pyrochemical processing of Idaho Chemical Processing Plant (ICPP) High Level Waste (HLW) calcine

    International Nuclear Information System (INIS)

    Bronson, M.C.; Ebbinghaus, B.B.; Riley, D.C.; Nelson, L.; Del Debbio, J.

    1994-01-01

    Inertial force damping control by micromanipulator modulation is proposed to suppress the vibrations of a micro/macro-manipulator system. The proposed controller, developed using classical control theory, is added to the existing control system. The proposed controller uses real-time measurements of macro-manipulator flexibility to adjust the motion of the micro manipulator to counteract structural vibrations. Experimental studies using an existing micro/macro flexible link manipulator testbed demonstrate the effectiveness of the proposed approach to suppression of vibrations in the macro/micro-manipulator system using micromanipulator-based inertial active damping control

  5. High-level radioactive waste disposal problem in Russia

    International Nuclear Information System (INIS)

    Velichkin, Vasily I.

    1999-01-01

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

  6. High-level waste repository-induced effects

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  7. High-level waste repository-induced effects

    International Nuclear Information System (INIS)

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

    2016-10-01

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

  8. Disposal of high level radioactive wastes in geological formations

    International Nuclear Information System (INIS)

    Martins, L.A.M.; Carvalho Bastos, J.P. de

    1978-01-01

    The disposal of high-activity radioactive wastes is the most serious problem for the nuclear industry. Among the solutions, the disposal of wastes in approriated geological formations is the most realistic and feasible. In this work the methods used for geological disposal, as well as, the criteria, programs and analysis for selecting a bite for waste disposal are presented [pt

  9. Proposed methods for treating high-level pyrochemical process wastes

    International Nuclear Information System (INIS)

    Johnson, T.R.; Miller, W.E.; Steunenberg, R.K.

    1985-01-01

    This survey illustrates the large variety and number of possible techniques available for treating pyrochemical wastes; there are undoubtedly other process types and many variations. The choice of a suitable process is complicated by the uncertainty as to what will be an acceptable waste form in the future for both TRU and non-TRU wastes

  10. 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

  11. Alternatives for long-term management of defense high-level radioactive waste, Hanford Reservations, Richland, Washington

    International Nuclear Information System (INIS)

    1977-09-01

    The objective of this document is to provide information or alternatives that are being considered for the long-term management of defense high-level radioactive waste stored at Hanford in underground tanks and in stainless steel-lined concrete basins. For purposes of basic programmatic decision making, four major alternatives based on disposal location are considered. The steps leading to placement of the waste in the following locations are illustrated: existing waste tanks; onsite engineered surface facilities; onsite geologic repository; and offsite geologic repository. The four major disposal alternatives are expanded into 27 alternative plans by considering: (1) variations in the final form of the high-level fraction (with radionuclide removal) to include glass, concrete, and powder; (2) variations in the final form of the dehydrated waste product to include glass, calcined clay, and powder; and (3) variations in the treatment and handling of encapsulated waste to include packaging of capsules in canisters and conversion of the strontium fluoride and cesium chloride to glass; canisters stored in sealed casks on the surface are disposed of in a surface vault after the radionuclides have decayed sufficiently to avoid a heat-transfer problem. A description of the technology, a preliminary risk assessment, and preliminary cost estimates for each of these 27 plans are presented. The technology required to implement any of the 27 alternative plans has not been developed to the point where any plan can be considered completely technically sound and feasible

  12. Risk comparison of different treatment and disposal strategies of high level liquid radioactive waste

    International Nuclear Information System (INIS)

    Fang Dong

    1997-01-01

    The risk of different treatment and disposal strategies of high level liquid radioactive waste from spent fuel reprocessing is estimated and compared. The conclusions obtained are that risk difference from these strategies is very small and high level liquid waste can be reduced to middle and low level waste, if the decontamination factor for 99 Tc is large enough, which is the largest risk contributor in the high level radioactive waste from spent fuel reprocessing. It is also shown that the risk of high level radioactive waste could be reduced by the technical strategy of combining partitioning and transmutation

  13. Radioactive Waste Management Research Program Plan for high-level waste: 1987

    International Nuclear Information System (INIS)

    1987-05-01

    This plan will identify and resolve technical and scientific issues involved in the NRC's licensing and regulation of disposal systems intended to isolate high level hazardous radioactive wastes (HLW) from the human environment. The plan describes the program goals, discusses the research approach to be used, lays out peer review procedures, discusses the history and development of the high level radioactive waste problem and the research effort to date and describes study objectives and research programs in the areas of materials and engineering, hydrology and geochemistry, and compliance assessment and modeling. The plan also details the cooperative interactions with international waste management research programs. Proposed Earth Science Seismotectonic Research Program plan for radioactive waste facilities is appended

  14. Mobile calcination and cementation unit for solidification of concentrated radioactive wastes

    International Nuclear Information System (INIS)

    Napravnik, J.; Sazavsky, P.; Skaba, V.; Skvarenina, R.; Ditl, P.

    1985-01-01

    Mobile experimental unit MESA-1 was developed and manufactured for processing radioactive concentrates by direct cementation. The unit is mainly designed for processing low-level liquid wastes from nuclear power plants and other nuclear installations, in which the level of radioactivity does not exceed 10 10 Bq/m 3 , the salt content of liquid solutions does not exceed 500 kg/m 3 and the maximum amount of boric acid is 130 kg/m 3 . The equipment is built into three modules which may be assembled and dismantled in a short time and transported separately. The unit without the calciner module was tested in non-radioactive mode and in operation with actual radioactive wastes from the V-1 nuclear power plant. The course and results of the tests are described in detail. All project design values were achieved, a total of 18 dm 3 model solutions were processed and 1 m 3 of actual wastes with a salt content of 450 kg/m 3 . The test showed that with regard to the radiation level reached it will be necessary in the process of calcination to increase the shielding of certain exposed points. The calciner module is being assembled for completion. (Z.M.)

  15. An alternative waste form for the final disposal of high-level radioactive waste (HLW) on the basis of a survey of solidification and final disposal of HLW

    International Nuclear Information System (INIS)

    Bauer, C.

    1982-01-01

    The dissertation comprises two separate parts. The first part presents the basic conditions and concepts of the process leading to the development of a waste form, such as:origin, composition and characteristics of the high-level radioactive waste; evaluation of the methods available for the final disposal of radioactive waste, especially the disposal in a geological formation, including the resulting consequences for the conditions of state in the surroundings of the waste package; essential option for the conception of a waste form and presentation of the waste forms developed and examined on an international level up to now. The second part describes the production of a waste form on TiO 2 basis, in which calcined radioactive waste particles in the submillimeter range are embedded in a rutile matrix. That waste form is produced by uniaxial pressure sintering in the temperature range of 1223 K to 1423 K and pressures between 5 MPa and 20 MPa. Microstructure, mechanical properties and leaching rates of the waste form are presented. Moreover, a method is explained allowing compacting of the rutile matrix and also integration of a wasteless overpack of titanium or TiO 2 into the waste form. (orig.) [de

  16. The defense waste processing facility: the final processing step for defense high-level waste disposal

    International Nuclear Information System (INIS)

    Cowan, S.P.; Sprecher, W.M.; Walton, R.D.

    1983-01-01

    The policy of the U.S. Department of Energy is to pursue an aggressive and credible waste management program that advocates final disposal of government generated (defense) high-level nuclear wastes in a manner consistent with environmental, health, and safety responsibilities and requirements. The Defense Waste Processing Facility (DWPF) is an essential component of the Department's program. It is the first project undertaken in the United States to immobilize government generated high-level nuclear wastes for geologic disposal. The DWPF will be built at the Department's Savannah River Plant near Aiken, South Carolina. When construction is complete in 1989, the DWPF will begin processing the high-level waste at the Savannah River Plant into a borosilicate glass form, a highly insoluble and non-dispersable product, in easily handled canisters. The immobilized waste will be stored on site followed by transportation to and disposal in a Federal repository. The focus of this paper is on the DWPF. The paper discusses issues which justify the project, summarizes its technical attributes, analyzes relevant environmental and insitutional factors, describes the management approach followed in transforming technical and other concepts into concrete and steel, and concludes with observations about the future role of the facility

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

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Steven E.

    2013-11-11

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

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

    International Nuclear Information System (INIS)

    Kelly, Steven E.

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Jones, K.E.; Salmon, R.

    1990-08-01

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

  20. Criticality safety of high-level tank waste

    International Nuclear Information System (INIS)

    Rogers, C.A.

    1995-01-01

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

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

    Science.gov (United States)

    Dukert, Joseph M.

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

  2. Test plan: Effects of phase separation on waste loading for high level waste glasses

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    2000-01-01

    As part of the Tanks Focus Area's (TFA) effort to increase waste loading for high-level waste (HLW) vitrification at various facilities in the Department of Energy (DOE) complex, the occurrence of phase separation in waste glasses spanning the Savannah River Site (SRS) and Idaho National Engineering and Environmental Laboratory (INEEL) composition ranges were studied during FY99. The type, extent, and impact of phase separation on glass durability for a series of HLW glasses, e.g., SRS-type and INEEL-type, were examined

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

    International Nuclear Information System (INIS)

    Richardson, G.L.

    1976-06-01

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

  4. High-level radioactive waste incorporation into (special) cements

    International Nuclear Information System (INIS)

    Roy, D.M.; Gouda, G.R.

    1978-01-01

    A feasibility study has demonstrated that very strong, durable, relatively impermeable cylinders may be prepared by hot pressing combinations of cements with simulated radioactive waste solids. While the properties have not been studied exhaustively, the results suggest an optional method for immobilization and isolation of radioactive waste. Samples prepared with calcium aluminate cements appeared to have properties superior to those with Portland cements. Four simulated radioactive waste compositions having high rare-earth oxide contents, and some containing a large excess of NaNO 3 , were studied. Modest temperatures [423 to 673 K (150 to 400 0 C)] were used for hot pressing at pressures from 178 to 345 MPa. Dense strong very low porosity specimens resulted when mixtures containing from 10 to 50% waste were hot pressed, incorporating also a small percentage of water. In addition, high-strength cement cylinders were prepared with the waste solid (approximately 20 wt% waste) in a separate core and were very resistant to leaching by water near its boiling point. With this configuration, even the NaNO 3 -containing wastes were resistant to leaching by water

  5. Geology of high-level nuclear waste disposal: an introduction

    International Nuclear Information System (INIS)

    Roxbugh, I.S.

    1987-01-01

    Hazardous waste is produced by the nuclear fuel cycle from mining and milling of uranium ore, refinement and enrichment, reactor use, and during reprocessing of spent fuel. Waste can be classified according to origin, physical state, and levels of radioactivity and radiotoxicity. The method of the long-term waste disposal is based on the degree of the hazard and the length of time (1000 years to millions of years) for the waste to become safe. The International Atomic Energy Agency (IAEA) has classified radioactive waste into five categories (I-V) based on the amount of radioactivity and heat output of the waste. The text is concerned mainly with the two most hazardous categories (I and II). Disposal at various geological sites using proven mining, engineering, and deep drilling techniques has been proposed and studied. An ideal geological repository would have (1) minimum ground water movement, (2) geochemical and mineralogical properties to retard or immobilize the effects of the nuclear waste from reaching the biosphere, (3) thermochemical properties to allow for heat loading without damage, and (4) structural strength for the operational period. Types of geological environments (both undersea and on land) include evaporites, crystalline rocks, and argillaceous deposits. European and North American case histories are described, and there is a glossary and an extensive list of references in this concise review

  6. Probabilistic safety assessment for Hanford high-level waste tanks

    International Nuclear Information System (INIS)

    MacFarlane, D.R.; Stack, D.S.; Kindinger, J.P.; Deremer, R.K.

    1995-01-01

    This paper gives results from the first comprehensive level-3 probabilistic safety assessment (PSA), including consideration of external events, for the Hanford tank farm (HTF). This work was sponsored by the U.S. Department of Energy/Environmental Restoration and Waste Management Division (DOE/EM). At the HTF, there are 177 underground tanks in 18 separate tank farms containing accumulated liquid/sludge/saltcake radioactive wastes from 50 yr of weapons materials production activities. The total waste volume is ∼60 million gal, containing ∼200 million Ci of radioactivity

  7. Processing of concentrated radioactive wastes into cement and bitumens following calcination

    International Nuclear Information System (INIS)

    Napravnik, J.; Sazavsky, P.; Ditl, P.; Prikryl, P.

    1985-01-01

    A brief characteristic is presented of the most frequently used processes of solidification of liquid radioactive wastes, viz., bituminization, cementation and their combination with calcination. The effect of individual parameters is assessed on the choice of the type of solidification process as is their importance in the actual process, in temporary storage, during transportation and under conditions of long-term storage. It has been found that a combination of the procedures could lead to a modular system of methods and equipment. This would allow to approach optimal solidification of wastes in the present period and to establish a research reserve for the development of more modern, economically advantageous and safer procedures. A rough estimate is made of the costs of the solidification of 1 m 3 of radioactive concentrate from the V-1 power plant at a production of 380 m 3 /year, this for the cementation-calcination and bituminization-calcination procedures. The said rough economic analysis only serves to identify the major operating components which have the greatest effect on the economic evaluation of the solidification procedures. (Z.M.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Marra, James [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kim, Dong -Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maio, Vincent [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-02

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these "troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approached to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.

  9. 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

  10. The management of high level waste and its environmental impact

    International Nuclear Information System (INIS)

    Saunders, P.A.H.

    1981-01-01

    This paper summarises the techniques that are used for the management of the radioactive wastes that result from the nuclear generation of electricity and that cannot be released directly into the environment. The quantities and characteristics of the wastes are outlined and a description is given of current and probable future stores and their environmental impact. The research and development programme that is being undertaken to establish the safety and environmental impact of an ultimate repository is discussed and a comparison is made between the activities and toxic potentials of the wastes and those of naturally occurring materials, fossil fuels and fertilisers. It is concluded that the wastes can be managed without undue risk to man or to the environment. (author)

  11. High-level waste solidification - why we chose glass

    International Nuclear Information System (INIS)

    Grover, J.R.

    1979-05-01

    This paper considers the desirable properties and factors to be assessed in the selection of a solidified waste product, surveys the possible product options and then analyses in detail their suitability in meeting the criteria. (author)

  12. New Technological Options to Manage High Level Waste

    International Nuclear Information System (INIS)

    Gonzalez Romero, E. M.

    2007-01-01

    Nuclear energy renaissance and its expansion in time and space has renewed the need for minimization technologies applicable to nuclear wastes. The minimization technologies include new power reactor concepts, Generation IV, and dedicated technologies like Partitioning and Transmutation of the actinides contained in the spent fuel. These technologies apply the principle of classification and recycling to the spent fuel to transform what at present is an environmental hazard into an energy source. the waste minimization technologies are also relevant for countries planning the reduction or phase-out of nuclear energy, as they will allow minimizing the size and number of the final waste repositories. Present estimations indicate that reductions as large as a factor 100 in the amount (radiotoxicity) of long lived nuclear waste are feasibly, with a modest increase on the final electricity cost. (Author)

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

    International Nuclear Information System (INIS)

    2005-01-01

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

  14. Liquid level measurement in high level nuclear waste slurries

    International Nuclear Information System (INIS)

    Weeks, G.E.; Heckendorn, F.M.; Postles, R.L.

    1990-01-01

    Accurate liquid level measurement has been a difficult problem to solve for the Defense Waste Processing Facility (DWPF). The nuclear waste sludge tends to plug or degrade most commercially available liquid-level measurement sensors. A liquid-level measurement system that meets demanding accuracy requirements for the DWPF has been developed. The system uses a pneumatic 1:1 pressure repeater as a sensor and a computerized error correction system. 2 figs

  15. Regulatable elements in the high-level waste management program

    International Nuclear Information System (INIS)

    Oakley, D.

    1979-01-01

    Regulatable elements of a deep geological nuclear waste isolation system are those characteristics of a candidate system which need to be specified to achieve control of its performance. This report identifies the regulatable elements with respect to waste form, repository design, site suitability, and the modeling and decision analysis processes. Regulatable elements in each section are listed and described briefly as they affect the short-term and long-term performance of a deep geological repository

  16. High-level waste disposal, ethics and thermodynamics

    Science.gov (United States)

    Schwartz, Michael O.

    2008-06-01

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

  17. Long-term high-level waste technology. Composite quarterly technical report, October-December 1979

    International Nuclear Information System (INIS)

    Cornman, W.R.

    1980-06-01

    This document summarizes work for the immobilization of high-level radioactive wastes from the chemical reprocessing of nuclear reactor fuels. The progress is reported in two main areas: site technology, and alternative waste form development

  18. An optimal retrieval, processing, and blending strategy for immobilization of Hanford high-level tank waste

    International Nuclear Information System (INIS)

    Hoza, M.

    1996-01-01

    Hanford tank waste will be separated into high-level and low-level portions; each portion will then be vitrified (other waste forms are also being considered for low-level waste) to produce a stable glass form for disposal. Because of the wide variability in the tank waste compositions, blending is being considered as a way to reduce the number of distinct compositions that must be vitrified and to minimize the resultant volume of vitrified waste. Three years of computational glass formulation and blending studies have demonstrated that blending of the high-level waste before vitrification can reduce the volume of high-level waste glass required by as much as 50 percent. This level of reduction would be obtained if all the high-level waste were blended together (Total Blend) prior to vitrification, requiring the retrieval and pretreatment of all tank waste before high-level vitrification was started. This paper will present an overall processing strategy that should be able to match the blending performance of the Total Blend and be more logistically feasible. The strategy includes retrieving, pretreating, blending and vitrifying Hanford tank waste. This strategy utilizes blending both before and after pretreatment. Similar wastes are blended before pretreatment, so as not to dilute species targeted for removal. The high-level portions of these pretreated early blends are then selectively blended to produce a small number of high-level vitrification feed streams

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

    Energy Technology Data Exchange (ETDEWEB)

    Marra, James [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kim, Dong -Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maio, Vincent [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these “troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (also with high Al2O3 concentrations). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group. An extended duration CCIM melter test was conducted on an AZ-101 waste simulant using the CCIM platform at the Idaho National Laboratory (INL). The melter was continually operated for approximately 80 hours demonstrating that the AZ-101 high waste loading glass composition could be readily processed using the CCIM technology. The resulting glass was close to the targeted composition and exhibited excellent durability in both

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

    International Nuclear Information System (INIS)

    Walker, D.D.

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    WILLIS, W.L.

    2000-01-01

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

  2. Corrosion issues in high-level nuclear waste containers

    Science.gov (United States)

    Asl, Samin Sharifi

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

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

    International Nuclear Information System (INIS)

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

    1989-11-01

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

  4. 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

  5. The vitrification of high level wastes using microwave power

    International Nuclear Information System (INIS)

    Hardwick, W.H.; Gayler, R.; Murphy, V.

    1981-01-01

    A process for radioactive waste vitrification which exploits advantages peculiar to microwave heating is under development. The advantages claimed are the removal of the heat source from the radioactive environment, the elimination of heat transfer barriers by direct coupling of the energy with the process materials, and the ability to evaporate liquors absorbed in a glass fibre matrix which constitutes the glass forming additive. This glass fibre matrix which constitutes the glass forming additive. This glass fibre is also used to filter off-gases and give a condensate free of solids. The fibre loaded with dried waste is converted to a homogeneous glass by melting using microwave power. (orig./DG)

  6. Effects of waste content of glass waste forms on Savannah River high-level waste disposal costs

    International Nuclear Information System (INIS)

    McDonell, W.R.; Jantzen, C.M.

    1985-01-01

    Effects of the waste content of glass waste forms of Savannah River high-level waste disposal costs are evaluated by their impact on the number of waste canisters produced. Changes in waste content affect onsite Defense Waste Processing Facility (DWPF) costs as well as offsite shipping and repository emplacement charges. A nominal 1% increase over the 28 wt % waste loading of DWPF glass would reduce disposal costs by about $50 million for Savannah River wastes generated to the year 2000. Waste form modifications under current study include adjustments of glass frit content to compensate for added salt decontamination residues and increased sludge loadings in the DWPF glass. Projected cost reductions demonstrate significant incentives for continued optimization of the glass waste loadings. 13 refs., 3 figs., 3 tabs

  7. The disposal of high-level radioactive waste. Vol. 2

    International Nuclear Information System (INIS)

    Parker, F.L.; Broshears, R.E.; Pasztor, J.

    1984-01-01

    This second volume contains the topics: activities of ICRP, UNSCEAR, IAEA, CEC, and NEA; review of natural programs; alternative concepts; responses to technical questions. The persons interviewed were mainly professionals responsible for nuclear waste management and regulation, but wherever possible, the views of other interested parties were sought, e.g. environmental or ecological groups. (HP)

  8. Sampling and analyses of SRP high-level waste sludges

    International Nuclear Information System (INIS)

    Stone, J.A.; Kelley, J.A.; McMillan, T.S.

    1976-08-01

    Twelve 3-liter samples of high-heat waste sludges were collected from four Savannah River Plant waste tanks with a hydraulically operated sample collector of unique design. Ten of these samples were processed in Savannah River Laboratory shielded cell facilities, yielding 5.3 kg of washed, dried sludge products for waste solidification studies. After initial drying, each batch was washed by settling and decantation to remove the bulk of soluble salts and then was redried. Additional washes were by filtration, followed by final drying. Conclusions from analyses of samples taken during the processing steps were: (a) the raw sludges contained approximately 80 wt percent soluble salts, most of which were removed by the washes; (b) 90 Sr and 238 , 239 Pu remained in the sludges, but most of the 137 Cs was removed by washing; (c) small amounts of sodium, sulfate, and 137 Cs remained in the sludges after thorough washing; (d) no significant differences were found in sludge samples taken from different risers of one waste tank. Chemical and radiometric compositions of the sludge product from each tank were determined. The sludges had diverse compositions, but iron, manganese, aluminum, and uranium were principal elements in each sludge. 90 Sr was the predominant radionuclide in each sludge product

  9. High level waste vitrification at the SRP [Savannah River Plant] (DWPF [Defense Waste Processing Facility] summary)

    International Nuclear Information System (INIS)

    Weisman, A.F.; Knight, J.R.; McIntosh, D.L.; Papouchado, L.M.

    1988-01-01

    The Savannah River Plant has been operating a nuclear fuel cycle since the early 1950's. Fuel and target elements are fabricated and irradiated to produce nuclear materials. After removal from the reactors, the fuel elements are processed to extract the products, and waste is stored. During the thirty years of operation including evaporation, about 30 million gallons of high level radioactive waste has accumulated. The Defense Waste Processing Facility (DWPF) under construction at Savannah River will process this waste into a borosilicate glass for long-term geologic disposal. The construction of the DWPF is about 70% complete; this paper will describe the status of the project, including design demonstrations, with an emphasis on the melter system. 9 figs

  10. Idaho National Engineering and Environmental Laboratory, Old Waste Calcining Facility, Scoville vicinity, Butte County, Idaho -- Photographs, written historical and descriptive data. Historical American engineering record

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    This report describes the history of the Old Waste Calcining Facility. It begins with introductory material on the Idaho National Engineering and Environmental Laboratory, the Materials Testing Reactor fuel cycle, and the Idaho Chemical Processing Plant. The report then describes management of the wastes from the processing plant in the following chapters: Converting liquid to solid wastes; Fluidized bed waste calcining process and the Waste Calcining Facility; Waste calcining campaigns; WCF gets a new source of heat; New Waste Calcining Facility; Last campaign; Deactivation and the RCRA cap; Significance/context of the old WCF. Appendices contain a photo key map for HAER photos, a vicinity map and neighborhood of the WCF, detailed description of the calcining process, and chronology of WCF campaigns.

  11. Idaho National Engineering and Environmental Laboratory, Old Waste Calcining Facility, Scoville vicinity, Butte County, Idaho -- Photographs, written historical and descriptive data. Historical American engineering record

    International Nuclear Information System (INIS)

    1997-01-01

    This report describes the history of the Old Waste Calcining Facility. It begins with introductory material on the Idaho National Engineering and Environmental Laboratory, the Materials Testing Reactor fuel cycle, and the Idaho Chemical Processing Plant. The report then describes management of the wastes from the processing plant in the following chapters: Converting liquid to solid wastes; Fluidized bed waste calcining process and the Waste Calcining Facility; Waste calcining campaigns; WCF gets a new source of heat; New Waste Calcining Facility; Last campaign; Deactivation and the RCRA cap; Significance/context of the old WCF. Appendices contain a photo key map for HAER photos, a vicinity map and neighborhood of the WCF, detailed description of the calcining process, and chronology of WCF campaigns

  12. The Defense Waste Processing Facility: an innovative process for high-level waste immobilization

    International Nuclear Information System (INIS)

    Cowan, S.P.

    1985-01-01

    The Defense Waste Processing Facility (DWPF), under construction at the Department of Energy's Savannah River Plant (SRP), will process defense high-level radioactive waste so that it can be disposed of safely. The DWPF will immobilize the high activity fraction of the waste in borosilicate glass cast in stainless steel canisters which can be handled, stored, transported and disposed of in a geologic repository. The low-activity fraction of the waste, which represents about 90% of the high-level waste HLW volume, will be decontaminated and disposed of on the SRP site. After decontamination the canister will be welded shut by an upset resistance welding technique. In this process a slightly oversized plug is pressed into the canister opening. At the same time a large current is passed through the canister and plug. The higher resistance of the canister/plug interface causes the heat which welds the plug in place. This process provides a high quality, reliable weld by a process easily operated remotely

  13. Modeling for speciation of radionuclides in waste packages with high-level radioactive wastes

    International Nuclear Information System (INIS)

    Weyand, Torben; Bracke, Guido; Seher, Holger

    2016-10-01

    Based on a literature search on radioactive waste inventories adequate thermodynamic data for model inventories were derived for geochemical model calculations using PHREEQC in order to determine the solid phase composition of high-level radioactive wastes in different containers. The calculations were performed for different model inventories (PWR-MOX, PWR-UO2, BWR-MOX, BMR-UO2) assuming intact containers under reduction conditions. The effect of a defect in the container on the solid phase composition was considered in variation calculations assuming air contact induced oxidation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2018-01-11

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

  15. Procedure for conditioning high-level solidified wastes

    Energy Technology Data Exchange (ETDEWEB)

    Hild, W; Krause, H; Scheffler, K

    1974-05-30

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

  16. Hanford high level waste: Sample Exchange/Evaluation (SEE) Program

    International Nuclear Information System (INIS)

    King, A.G.

    1994-08-01

    The Pacific Northwest Laboratory (PNL)/Analytical Chemistry Laboratory (ACL) and the Westinghouse Hanford Company (WHC)/Process Analytical Laboratory (PAL) provide analytical support services to various environmental restoration and waste management projects/programs at Hanford. In response to a US Department of Energy -- Richland Field Office (DOE-RL) audit, which questioned the comparability of analytical methods employed at each laboratory, the Sample Exchange/Exchange (SEE) program was initiated. The SEE Program is a selfassessment program designed to compare analytical methods of the PAL and ACL laboratories using sitespecific waste material. The SEE program is managed by a collaborative, the Quality Assurance Triad (Triad). Triad membership is made up of representatives from the WHC/PAL, PNL/ACL, and WHC Hanford Analytical Services Management (HASM) organizations. The Triad works together to design/evaluate/implement each phase of the SEE Program

  17. The disposal of high level nuclear waste in the oceans

    International Nuclear Information System (INIS)

    Vilks, Gustavs

    1976-01-01

    A report is given on a meeting held at Woods Hole, Massachusetts to consider the feasibility of using the sea bed as a disposal site for highly radioactive nuclear waste. Some disadvantages are explained, the chief being that ocean water alone, regardless of depth, is a poor barrier. Some delegates discussed emplacement of vitrified waste on the ocean floor, and others its burial in rock or sediment below the floor. The most suitable sites are the mid-plate/mid-gyse localities of abyssal hills. Some engineering work on submarine burial has been done by Sandia Labs., in the U.S.A. Ocean disposal is particularly interesting to Britain and Japan. Data on biological transport rates are needed. (author)

  18. F/H area high level waste tank status report

    International Nuclear Information System (INIS)

    Hayes, C.R. Jr.; Wells, M.N.

    1997-03-01

    Section IX.E.3 of the SRS Federal Facility Agreement requires the DOE to submit to EPA and SCDHEC, an annual report on the status of tanks being removed from service. Tanks that are slated for removal from service either do not meet secondary containment standards or have leak sites. The attached document is intended to meet this annual report requirement. An updated status of relevant portions of the Waste Removal Plan and Schedule is also included

  19. High-level waste processing and conditioning: vitrification

    International Nuclear Information System (INIS)

    Bonniaud, R.

    1981-02-01

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

  20. High-level waste solidification: why we chose glass

    International Nuclear Information System (INIS)

    Grover, J.R.

    1980-01-01

    This paper considers the desirable properties and factors to be assessed in the selection of a solidified waste product, surveys the possible product options and then analyzes in detail their suitability in meeting the criteria. It concludes that glasses are currently the preferred choice for the following reasons: their ability to fix the full spectrum of elements contained in the waste; their tolerance of the composition variations that will occur on a day to day basis in practice; their relatively low formation temperatures that lead to simpler and hence safer processing; their radiation stability; and their adequate leach rates. Suitable compositions are available for the wastes that will arise in the UK and techniques are available for manufacture on a production scale. Lower leach rates might be obtained by choosing glasses with higher formation temperatures or ceramics. However, these latter generally also have higher formation temperatures, have less tolerance for composition variations and their radiation stability is unproven. Supercalcines and synthetic rocks (SYNROC) may eventually be demonstrated to have some advantageous properties, but present indications are that these could be major disadvantages which more than offset any gains. Other advanced concepts (for example, the dispersion of glass beads in a metal matrix) have lower leach rates, but lead to additional complexity in manufacture

  1. Midwestern High-Level Radioactive Waste Transportation Project

    International Nuclear Information System (INIS)

    Sattler, L.R.

    1992-02-01

    In addition to arranging for storage and disposal of radioactive waste, the US Department of Energy (DOE) must develop a safe and efficient transportation system in order to deliver the material that has accumulated at various sites throughout the country. The ability to transport radioactive waste safely has been demonstrated during the past 20 years: DOE has made over 2,000 shipments of spent fuel and other wastes without any fatalities or environmental damage related to the radioactive nature of the cargo. To guarantee the efficiency of the transportation system, DOE must determine the optimal combination of rail transport (which allows greater payloads but requires special facilities) and truck transport Utilizing trucks, in turn, calls for decisions as to when to use legal weight trucks or, if feasible, overweight trucks for fewer but larger shipments. As part of the transportation system, the Facility Interface Capability Assessment (FICA) study contributes to DOE's development of transportation plans for specific facilities. This study evaluates the ability of different facilities to receive, load and ship the special casks in which radioactive materials will be housed during transport In addition, the DOE's Near-Site Transportation Infrastructure (NSTI) study (forthcoming) will evaluate the rail, road and barge access to 76 reactor sites from which DOE is obligated to begin accepting spent fuel in 1998. The NSTI study will also assess the existing capabilities of each transportation mode and route, including the potential for upgrade

  2. 3-D Mapping Technologies For High Level Waste Tanks

    International Nuclear Information System (INIS)

    Marzolf, A.; Folsom, M.

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    Manaktala, H.K.; Interrante, C.G.

    1990-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Manaktala, H.K. (Southwest Research Inst., San Antonio, TX (USA). Center for Nuclear Waste Regulatory Analyses); Interrante, C.G. (Nuclear Regulatory Commission, Washington, DC (USA). Div. of High-Level Waste Management)

    1990-12-01

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

  5. Simulation and characterization of a Hanford high-level waste slurry

    International Nuclear Information System (INIS)

    Russell, R.L.; Smith, H.D.

    1996-09-01

    The baseline waste used for this simulant is a blend of wastes from tanks 101-AZ, 102-AZ, 106-C, and 102-AY that have been through water washing. However, the simulant used in this study represents a combination of tank waste slurries and should be viewed as an example of the slurries that might be produced by blending waste from various tanks. It does not imply that this is representative of the actual waste that will be delivered to the privatization contractor(s). This blended waste sludge simulant was analyzed for grain size distribution, theological properties both as a function of concentration and aging, and calcining characteristics. The grain size distribution allows a comparison with actual waste with respect to theological properties. Slurries with similar grain size distributions of the same phases are expected to exhibit similar theological properties. Rheological properties may also change because of changes in the slurry's particulate supernate chemistry due to aging. Low temperature calcination allows the potential for hazardous gas generation to be investigated

  6. Volume reduction of low- and medium-level waste by incineration/calcination

    International Nuclear Information System (INIS)

    Buzonniere, A. de; Gauthey, J.C.

    1993-01-01

    Nuclear installations generate large quantities of low- and medium-level radwaste. This waste comes from various installations in the fuel cycle, reactor operation, research institute, hospitals, nuclear plate dismantling, etc.. TECHNICATOME did the project development work for the incineration plant of PIERRELATE (France) on behalf of COGEMA (Compagnie Generale des d'Etudes Technique). This plant has been in active service since November 1987. In addition, TECHNICATOME was in charge of the incinerator by a turnkey contract. This incinerator was commissioned in 1992. For a number of years, TECHNICATOME has been examining, developing and producing incineration and drying/calcination installations. They are used for precessing low- and medium-level radwaste

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

    International Nuclear Information System (INIS)

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

    1993-09-01

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

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

    International Nuclear Information System (INIS)

    Decamps, F.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1977-02-01

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

  10. Testing and evaluation of solidified high-level waste forms

    International Nuclear Information System (INIS)

    De Batist Al, R.

    1983-01-01

    In addition to the preceding programme of the European Atomic Energy Community two new borosilicate glass compositions have been introduced. The chemical stability of these waste forms, in particular with respect to geological disposal conditions, is examined as well as effects of alpha-radiation and of devitrification. Leaching studies include theoretical and experimental investigations of the basic leaching mechanisms, the measurement of the leach rates of a number of critical radioisotopes and the influence on the leach rate of various parameters such as temperature, pressure pH and duration. Of particular interest is the simulation of repository conditions. Prelimimary results are described related to various mineral waters, granite and salt solutions. The surface layers generated on the waste forms during corrosion are investigated in detail using various experimental techniques such as scanning electron microscopy, X-ray analysis and alpha particle energy loss spectra measurements. The radiation stability was further tested by continuing investigations of the samples doped with 238 Pu in the course of the previous programme; density and leach rate variations were measured. Effects on the leach rate of devitrification resulting from various heat treatments of active glass samples were also investigated

  11. Testing and evaluation of solidified high-level waste forms

    International Nuclear Information System (INIS)

    Engelmann, C.

    1984-01-01

    The report describes research by several laboratories on the behaviour, in aqueous and salt environments, of borosilicate glass ceramics proposed for the solidification of nuclear wastes by the European Community. Results were obtained on inactive simulates, doped materials, and on borosilicate glass containing real radioactive waste. The influence of many important parameters were studied: leaching mode, nature of the leachant, pH, pressure, temperature, duration of the treatment, etc. The results of tests lasting for as little as a few hours or for as long as several hundred days, at temperatures up to 200 0 C or under pressures up to 200 bars, are presented. Numerous analytical techniques (ESCA, EMP, IRR, SEM, etc.) were used to determine the structure and the chemical composition of the altered layer developed by hydration at the glass surface. Information is also given on physical properties of the borosilicate glass: crystallization phase separation, alpha-irradiation stability, mechanical and thermal stability, etc. Finally, preliminary results on the structure and composition of hollandite ceramics are given

  12. High-level waste vitrification off-gas cleanup technology

    International Nuclear Information System (INIS)

    Hanson, M.S.

    1980-01-01

    This brief overview is intended to be a basis for discussion of needs and problems existing in the off-gas clean-up technology. A variety of types of waste form and processes are being developed in the United States and abroad. A description of many of the processes can be found in the Technical Alternative Documents (TAD). Concurrently, off-gas processing systems are being developed with most of the processes. An extensive review of methodology as well as decontamination factors can be found in the literature. Since it is generally agreed that the most advanced solidification process is vitrification, discussion here centers about the off-gas problems related to vitrification. With a number of waste soldification facilities around the world in operation, it can be shown that present technology can satisfy the present requirement for off-gas control. However, a number of areas within the technology base show potential for improvement. Fundamental as well as verification studies are needed to obtain the improvements

  13. Mineralogic studies of tuff for high-level waste disposal

    International Nuclear Information System (INIS)

    Vaniman, D.; Bish, D.; Broxton, D.; Byers, F.; Carlos, B.; Levy, S.

    1986-01-01

    The volcanic rocks at Yucca Mountain, Nevada, consist predominantly of tuff that originated 12 to 14 million years ago as flows and airfalls of hot volcanic particulates. On cooling these units formed two major rock types: crystallized zones formed mostly of feldspar and silica minerals, and zones of glass. Alteration of glass to zeolite minerals occurred largely during structural tilting of Yucca Mountain in the ∼1-3 million years following the major eruptions. The compositions of zeolites formed from glasses strongly indicate open-system chemical exchange. Superimposed on this general alteration of glasses are areas of local high-temperature alteration. High-temperature alteration ended by 11 million years ago. Zeolites such as clinoptilolite persisted during high-temperature alteration at temperatures up to 100 degree C, suggesting that clinoptilolite at Yucca Mountain close to the thermally disturbed zone around a repository may also survive heating to temperatures at least this high. The mineralogic data from tuff at Yucca Mountain will ultimately be used by the Department of Energy Nevada Nuclear Waste storage Investigations for (1) defining the mineralogic component in estimating waste element travel times away from the repository and (2) determining the past history of alteration and the anticipated stability of minerals near the repository

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  15. High-level waste characterization at West Valley: Progress report for the period 1982-1985

    International Nuclear Information System (INIS)

    Rykken, L.E.

    1986-01-01

    This is a report on the work that was carried out at West Valley under the Waste Characterization Program. This Program covered a number of tasks in support of the design of facilities for the pretreatment and final encapsulation of the high level waste stored at West Valley. In particular, necessary physical, chemical, and radiological characterization of high-level reprocessing waste stored in two vaulted underground tanks was carried out over the period 1982 to 1985. 21 refs., 77 figs., 28 tabs

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

    International Nuclear Information System (INIS)

    CALMUS, R.B.

    1999-01-01

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

  17. Calcine production and management

    International Nuclear Information System (INIS)

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

    1979-01-01

    The process technology related to calcination of power reactor wastes is summarized. The primary calcination processes developed are spray calcination, fluidized-bed calcination, and rotary kiln calcination. Calcines from the spray calciner and rotary kiln are fed directly to a glassification process. The fluidized-bed product can either be fed to a waste form conversion process or stored. The major process steps for calcinations are feed preparation, calcination and product handling, and off-gas cleanup. Feed systems for the three processes are basically similar. Gravity flow and pump pressurized systems have been used successfully. The major problems are fatigue failure of feed valve bellows, plugging by undissolved solids, and calibration of flowmeters. Process heat input is by electrical resistance heating for the spray and rotary kiln calciners and in-bed combustion or in-bed heat exchange for the fluidized-bed system. Low-melting solids which can cause scaling or solids agglomeration in any of the processes is a major calcination problem; however, feed blending, process operating conditions, and equipment design have successfully controlled solids agglomeration. Primary off-gas cleanup devices for particulates are cyclones, sintered metal filters, venturi scrubbers, and HEPA filters. Scrubbers, condensers, and solid adsorbents are used successfully for volatile ruthenium removal. The years of pilot-plant and plant-scale calcination testing and operation of the three systems have shown that reactor wastes can be calcined safely and practically. 11 figures, 2 tables

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  19. Recovery of Cs from high level radioactive waste

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  20. Surrounding rock stress analysis of underground high level waste repository

    International Nuclear Information System (INIS)

    Liu Wengang; Wang Ju; Wang Guangdi

    2006-01-01

    During decay of nuclear waste, enormous energy was released, which results in temperature change of surrounding rock of depository. Thermal stress was produced because thermal expansion of rock was controlled. Internal structure of surrounding rock was damaged and strength of rock was weakened. So, variation of stress was a dynamic process with the variation of temperature. BeiShan region of Gansu province was determined to be the depository field in the future, it is essential to make research on granite in this region. In the process of experiment, basic physical parameters of granite were analyzed preliminary with MTS. Long range temperature and stress filed was simulated considering the damage effect of surrounding rock, and rules of temperature and stress was achieved. (authors)

  1. Transportation of high-level waste and spent fuel

    International Nuclear Information System (INIS)

    Carlson, J.H.; Lake, W.H.; Thompson, J.H.

    1993-01-01

    The Office of Civilian Radioactive Waste Management (OCRWM) transportation program is a multifaceted undertaking to transport spent nuclear fuel from commercial reactors to temporary and permanent storage facilities commencing in 1998. One of the significant ingredients necessary to achieving this goal is the development and acquisition of shipping casks. Efforts to design and acquire high capacity casks is ongoing, as are efforts to purchase casks that can be made available using current technology. By designing casks that are optimized to the specifications of the older cooler spent fuel that will be shipped, and by designing to current NRC requirements, OCRWM's new generation of spent fuel casks will be more efficient and at least as safe as current cask designs. (J.P.N.)

  2. Physical, Chemical and Structural Evolution of Zeolite - Containing Waste Forms Produced from Metakaolinite and Calcined HLW

    International Nuclear Information System (INIS)

    Grutzeck, Michael

    2005-01-01

    During the seventh year of the current grant (DE-FG02-05ER63966) we completed an exhaustive study of cold calcination and began work on the development of tank fill materials to fill empty tanks and control residuals. Cold calcination of low and high NOx low activity waste (LAW) SRS Tank 44 and Hanford AN-107 simulants, respectively with metallic Al + Si powders was evaluated. It was found that a combination of Al and Si powders could be used as reducing agents to reduce the nitrate and nitrite content of both low and high NOx LAW to low enough levels to allow the LAW to be solidified directly by mixing it with metakaolin and allowing it to cure at 90 C. During room temperature reactions, NOx was reduced and nitrogen was emitted as N2 or NH3. This was an important finding because now one can pretreat LAW at ambient temperatures which provides a low-temperature alternative to thermal calcination. The significant advantage of using Al and Si metals for denitration/denitrition of the LAW is the fact that the supernate could potentially be treated in situ in the waste tanks themselves. Tank fill materials based upon a hydroceramic binder have been formulated from mixtures of metakaolinite, Class F fly ash and Class C flue gas desulphurization (FGD) ash mixed with various concentrations of NaOH solution. These harden over a period of hours or days depending on composition. A systematic study of properties of the tank fill materials (leachability) and ability to adsorb and hold residuals is under way

  3. US program for the immobilization of high-level nuclear wastes

    International Nuclear Information System (INIS)

    Crandall, J.L.

    1979-01-01

    A program has been developed for long-term management of high-level nuclear waste. The Savannah River Operations Office of the US Department of Energy is acting as the lead office for this program with technical advice from the E.I. du Pont de Nemours and Company. The purpose of the long-term program is to immobilize the DOE high-level waste in forms that act as highly efficient barriers against radionuclide release to the disposal site and to provide technology for similar treatment of commercial high-level waste in case reprocessing of commercial nuclear fuels is ever resumed. Descriptions of existing DOE and commercial wastes, program strategy, program expenditures, development of waste forms, evaluation and selection of waste forms, regulatory aspects of waste form selection, project schedules, and cost estimates for immobilization facilities are discussed

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-04

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

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

    International Nuclear Information System (INIS)

    1977-04-01

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

  6. Plutonium Solubility In High-Level Waste Alkali Borosilicate Glass

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  7. Management of high level nuclear waste - the nordic approach

    International Nuclear Information System (INIS)

    Engstrom, S.; Aikas, T.

    2000-01-01

    Both the Swedish and the Finnish nuclear waste programmes are aimed at disposal of encapsulated spent nuclear fuel into the crystalline bedrock. In both countries research and development work have been performed since the 1970's. The focus of the programme in both countries is now shifting to practical demonstration of encapsulation technology. In parallel a site-selection programme is being carried out. Finland has selected a site at Eurajoki and is currently waiting for the Government to agree to the choice of the site. In Sweden, at least two sites will be selected by year 2001 with the goal, after performed drillings, to select one of them around 2008. Site selection for the deep repository is probably the most difficult and most sensitive part of the whole programme. The repository will be sited at a suitable place in Sweden respectively Finland where high safety requirements will be met with the consent of the concerned municipality. If there is a Nordic approach to tackle this issue that would probably be: - A stepwise approach in which the disposal is implemented in gradually each step having a decision making stage leading to a commitment of various parties involved to the following stage. -A total transparency of the work performed and the decision making process. - A genuine will from the industry to establish a dialogue with the public in the involved communities. - A will to take the time and the patience necessary to establish a constructive working relationship with the communities participating in the site selection. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-10-01

    Los Alamos National Laboratory has performed a comprehensive probabilistic safety assessment (PSA) that includes consideration of external events for the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases (``burps``) a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is underway in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. This PSA for Tank 101-SY, which did not consider the pump experiment or future tank-remediation activities, involved three distinct tasks. First, the accident sequence analysis identified and quantified those potential accidents whose consequences result in tank material release. Second, characteristics and release paths for the airborne and liquid radioactive source terms were determined. Finally, the consequences, primarily onsite and offsite potential health effects resulting from radionuclide release, were estimated, and overall risk curves were constructed. An overview of each of these tasks and a summary of the overall results of the analysis are presented in the following sections.

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

    International Nuclear Information System (INIS)

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

    1993-01-01

    Los Alamos National Laboratory has performed a comprehensive probabilistic safety assessment (PSA) that includes consideration of external events for the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases (''burps'') a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is underway in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. This PSA for Tank 101-SY, which did not consider the pump experiment or future tank-remediation activities, involved three distinct tasks. First, the accident sequence analysis identified and quantified those potential accidents whose consequences result in tank material release. Second, characteristics and release paths for the airborne and liquid radioactive source terms were determined. Finally, the consequences, primarily onsite and offsite potential health effects resulting from radionuclide release, were estimated, and overall risk curves were constructed. An overview of each of these tasks and a summary of the overall results of the analysis are presented in the following sections

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  11. Long-term stability of high-level waste forms

    International Nuclear Information System (INIS)

    Vernaz, E.; Loida, A.; Malow, G.; Marples, J.A.C.; Matzke, H.J.

    1990-01-01

    The long-term stability of HLW forms is reviewed with regard to temperature, irradiation and aqueous corrosion in a geological environment. The paper focuses on borosilicate glasses, but the radiation stability results are compared with some HLW ceramics. Thermal stability: most nuclear waste glass compositions have been adjusted to ensure a low final crystallized fraction. The crystallization of highly active Pamela glass samples was similar to that of nonradioactive glass. Radiation stability: No adverse effect of irradiation damage was found in glasses doped with short-lived actinides: volume changes were small, no significant change in the leach rate was observed, and the fracture toughness increased. For most ceramics investigated, volume changes of up to 9%, amorphization and higher leach rates were observed as a consequence of high α decay doses. For the KAB 78 ceramic, however, none of these effects were detected since the matrix was not subject to α recoil damage. Chemical stability: It has been demonstrated that alteration by water depends largely on the repository conditions. Most clay act as silica sinks, and increase the glass corrosion rate. It is possible, however, to specify realistic temperature, pressure and environmental conditions to ensure glass integrity for more than 10 000 years

  12. Removal of americium and curium from high-level wastes

    International Nuclear Information System (INIS)

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

    1979-01-01

    The DHDECMP process was demonstrated to give a 99.5% removal of actinides from actual HLLW in small-scale, batch extraction tests. Results from cold tests indicate that it may be possible to carry out the oxalate precipitation step of the OPIX process continuously. About 90% recovery of the trivalent actinides and lanthanides can be achieved in the continuous precipitation. the presence of zirconium impurity in feed solutions to Talspeak process at concentrations of 10 -4 M (0.5% of the Zr in the original waste) affected phase separations but equipment could be operated satisfactory in cold tests. Zirconium concentrations of 10 -3 M seriously affected phase separations and substantial quantities of interfacial cruds were formed. Modest concentrations (0.006 M or less) of H 2 MEHP, a suspected degradation product of HDEHP, did not effect separation factors. The presence of impurities derived from the thermal degradation of DHDECMP did not inhibit the loading of the trivalent actinide and lanthanide elements in the cation exchange chromatographic process for their separation. It appears that the biodentate (DHDECMP) solvent extraction process and the OPIX process are the leading candidate process for the co-removal of trivalent actinide and lanthanide elements from HLLW. The cation exchange chromatography and the Talspeak processes, are the leading candidate processes for the subsequent separation of actinides and lanthanides. The bidentate and cation exchange processes are further along in their development than the other processes and are currently considered the reference processes for the partitioning of Am-Cm from HLLW. 4 figures, 4 tables

  13. Hanford long-term high-level waste management program overview

    International Nuclear Information System (INIS)

    Reep, I.E.

    1978-05-01

    The objective is the long-term disposition of the defense high-level radioactive waste which will remain upon completion of the interim waste management program in the mid-1980s, plus any additional high-level defense waste resulting from the future operation of N Reactor and the Purex Plant. The high-level radioactive waste which will exist in the mid-1980s and is addressed by this plan consists of approximately 3,300,000 ft 3 of damp salt cake stored in single-shell and double-shell waste tanks, 1,500,000 ft 3 of damp sludge stored in single-shell and double-shell waste tanks, 11,000,000 gallons of residual liquor stored in double-shell waste tanks, 3,000,000 gallons of liquid wastes stored in double-shell waste tanks awaiting solidification, and 2,900 capsules of 90 SR and 137 Cs compounds stored in water basins. Final quantities of waste may be 5 to 10% greater, depending on the future operation of N Reactor and the Purex Plant and the application of waste treatment techniques currently under study to reduce the inventory of residual liquor. In this report, the high-level radioactive waste addressed by this plan is briefly described, the major alternatives and strategies for long-term waste management are discussed, and a description of the long-term high-level waste management program is presented. Separate plans are being prepared for the long-term management of radioactive wastes which exist in other forms. 14 figures

  14. Preconceptual design study for solidifying high-level waste: West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Hill, O.F.

    1981-04-01

    This report presents a preconceptual design study for processing radioactive high-level liquid waste presently stored in underground tanks at Western New York Nuclear Service Center (WNYNSC) near West Valley, New York, and for incorporating the radionculides in that waste into a solid. The high-level liquid waste accumulated from the operation of a chemical reprocessing plant by the Nuclear Fuel Services, Inc. from 1966 to 1972. The high-level liquid waste consists of approximately 560,000 gallons of alkaline waste from Purex process operations and 12,000 gallons of acidic (nitric acid) waste from one campaign of processing thoria fuels by a modified Thorex process (during this campaign thorium was left in the waste). The alkaline waste contains approximately 30 million curies and the acidic waste contains approximately 2.5 million curies. The reference process described in this report is concerned only with chemically processing the high-level liquid waste to remove radionuclides from the alkaline supernate and converting the radionuclide-containing nonsalt components in the waste into a borosilicate glass

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

    International Nuclear Information System (INIS)

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

    1998-09-01

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

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

    International Nuclear Information System (INIS)

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

    1994-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-05-01

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

  18. Disposal of high level nuclear wastes: Thermodynamic equilibrium and environment ethics

    Institute of Scientific and Technical Information of China (English)

    RANA Mukhtar Ahmed

    2009-01-01

    Contamination of soil, water or air, due to a failure of containment or disposal of high level nuclear wastes, can potentially cause serious hazards to the environment or human health. Essential elements of the environment and radioactivity dangers to it are illustrated. Issues of high level nuclear waste disposal are discussed with a focus on thermodynamic equilibrium and environment ethics. Major aspects of the issues are analyzed and described briefly to build a perception of risks involved and ethical implications. Nuclear waste containment repository should be as close as possible to thermodynamic equilibrium. A clear demonstration about safety aspects of nuclear waste management is required in gaining public and political confidence in any possible scheme of permanent disposal. Disposal of high level nuclear waste offers a spectrum of environment connected challenges and a long term future of nuclear power depends on the environment friendly solution of the problem of nuclear wastes.

  19. Interfaces between transport and geological disposal systems for high level radioactive waste and spent nuclear fuel

    International Nuclear Information System (INIS)

    1994-09-01

    This document is an IAEA publication which identifies and discusses the interfaces and the interface requirements between high level waste, the waste transport system used for carriage of the waste to the disposal facility, and the high level waste disposal facility. The development of this document was prompted in part by the initiatives in various Member States to select, characterize and design the facilities for potential high level waste geological repositories. These initiatives have progressed to the point where an international document would be useful in calling attention to the need for establishing, in a systematic way, interfaces and interface requirements between the transport systems to be used and the waste disposal packages and geological repository. Refs, figs and tabs

  20. Disposal of high level nuclear wastes: thermodynamic equilibrium and environment ethics

    International Nuclear Information System (INIS)

    Rana, M.A.

    2009-01-01

    Contamination of soil, water or air, due to a failure of containment or disposal of high level nuclear wastes, can potentially cause serious hazards to the environment or human health. Essential elements of the environment and radioactivity dangers to it are illustrated. Issues of high level nuclear waste disposal are discussed with a focus on thermodynamic equilibrium and environment ethics. Major aspects of the issues are analyzed and described briefly to build a perception of risks involved and ethical implications. Nuclear waste containment repository should be as close as possible to thermodynamic equilibrium. A clear demonstration about safety aspects of nuclear waste management is required in gaining public and political confidence in any possible scheme of permanent disposal. Disposal of high level nuclear waste offers a spectrum of environment connected challenges and a long term future of nuclear power depends on the environment friendly solution of the problem of nuclear wastes. (authors)

  1. OCRWM [Office of Civilian Radioactive Waste Management] publications catalog on high-level radioactive waste management

    International Nuclear Information System (INIS)

    1987-07-01

    The US Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM) is publishing this catalog to provide citations of selected technical and public information on the subject of high-level radioactive waste management. The catalog is a resource and reference tool. It will be updated and printed annually. The online catalog will be available for review through OCRWM's Product Record System (PRS) and can eventually be made available to the public. The printed catalog version is suitable for libraries and those individuals needing either a broad base of information or a particular source; the computerized catalog version provides the most current information resources available, since updates to citations will be made as they are received. The number of documents suitable for listing in this catalog is expected to grow significantly each year

  2. Efficient handling of high-level radioactive cell waste in a vitrification facility analytical laboratory

    International Nuclear Information System (INIS)

    Roberts, D.W.; Collins, K.J.

    1998-01-01

    The Savannah River Site''s (SRS) Defense Waste Processing Facility (DWPF) near Aiken, South Carolina, is the world''s largest and the United State''s first high level waste vitrification facility. For the past 1.5 years, DWPF has been vitrifying high level radioactive liquid waste left over from the Cold War. The vitrification process involves the stabilization of high level radioactive liquid waste into borosilicate glass. The glass is contained in stainless steel canisters. DWPF has filled more than 200 canisters 3.05 meters (10 feet) long and 0.61 meters (2 foot) diameter. Since operations began at DWPF in March of 1996, high level radioactive solid waste continues to be generated due to operating the facility''s analytical laboratory. The waste is referred to as cell waste and is routinely removed from the analytical laboratories. Through facility design, engineering controls, and administrative controls, DWPF has established efficient methods of handling the high level waste generated in its laboratory facility. These methods have resulted in the prevention of undue radiation exposure, wasted man-hours, expenses due to waste disposal, and the spread of contamination. This level of efficiency was not reached overnight, but it involved the collaboration of Radiological Control Operations and Laboratory personnel working together to devise methods that best benefited the facility. This paper discusses the methods that have been incorporated at DWPF for the handling of cell waste. The objective of this paper is to provide insight to good radiological and safety practices that were incorporated to handle high level radioactive waste in a laboratory setting

  3. Some legal aspects on high level radioactive waste disposal in Japan

    International Nuclear Information System (INIS)

    Tanabe, Tomoyuki

    1997-01-01

    In Japan, it is considered to be an urgent problem to prepare the system for the research and execution of high level radioactive waste disposal. Under what regulation scheme the disposal should be done has not been sufficiently examined. In this research, the examination was carried out on the legal aspects of high level radioactive waste disposal as follows. First, the current legislation on the disposal in Japan was analyzed, and it was made clear that high level radioactive waste disposal has not been stipulated clearly. Next, on the legal choices which are conceivable on the way the legislation for high level radioactive waste disposal should be, from the aspects of applying the law on regulating nuclear reactors and others, applying the law on nuclear power damage reparation, and industrialization by changing the government ordinances, those were arranged in six choices, and the examination was carried out for each choice from the viewpoints of the relation with the base stipulation for waste-burying business, the speciality of high level radioactive waste disposal as compared with other actions of nuclear power business, the coordination with existing nuclear power of nuclear power business, the coordination with existing nuclear power law system and the formation of national consensus. In this research, it was shown that the execution of high level radioactive waste disposal as the business based on the separate legislation is the realistic choice. (K.I.)

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

    International Nuclear Information System (INIS)

    Dupree, S.A.

    1977-06-01

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

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

    International Nuclear Information System (INIS)

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

    1985-02-01

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

  6. Status of commercial nuclear high-level waste disposal. Special report

    International Nuclear Information System (INIS)

    Dau, G.J.; Williams, R.F.

    1976-09-01

    The results of this review, presented in the form of a functional description of high level waste management system, shows that technology is available to dispose of nuclear waste safely by several different processes. The most attractive alternative in terms of available technology and shortness of time to demonstrate it at commercial scale is a system that converts the waste to a solid by immobilizing the radioactive elements in a glass matrix. Brief comments are also given on international efforts in high level waste management and advanced disposal concepts

  7. HIGH ALUMINUM HLW (HIGH LEVEL WASTE) GLASSES FOR HANFORD'S WTP (WASTE TREATMENT PROJECT)

    International Nuclear Information System (INIS)

    Kruger, A.A.; Bowan, B.W.; Joseph, I.; Gan, H.; Kot, W.K.; Matlack, K.S.; Pegg, I.L.

    2010-01-01

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

  8. Final technical position on documentation of computer codes for high-level waste management

    International Nuclear Information System (INIS)

    Silling, S.A.

    1983-06-01

    Guidance is given for the content of documentation of computer codes which are used in support of a license application for high-level waste disposal. The guidelines cover theoretical basis, programming, and instructions for use of the code

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Piccolo, S.F.

    1999-01-01

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

  11. Evaluation of S-type fiberglass composites for use in high-level radioactive waste environments

    International Nuclear Information System (INIS)

    Parra, S.A.

    1996-01-01

    Two types of S-type fiberglass materials were evaluated for use in a high-level radioactive waste environment. The S-type fiberglass composites tested were in the form of tubes and were exposed to a simulated high-level radioactive waste environment consisting of corrosive chemicals, high gamma radiation, and elevated temperatures. The physical properties of the exposed and unexposed tube samples were compared to determine the effects of the simulated environment on the S-type fiberglass composites

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

    International Nuclear Information System (INIS)

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

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

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

    International Nuclear Information System (INIS)

    Neumann, W.

    1980-01-01

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

  14. Neotectonic movement feature in preselection area for high level radioactive waste repository

    International Nuclear Information System (INIS)

    Huang Xianfang; Gao Yang; He Jianguo; Li Jianzhong; Gao Honglei; Xu Guoqing

    2010-01-01

    Neotectonic activity intensity is an important criteria for evaluating high level radioactive waste repository. The guiding ideology, methods and application of neotectonic study are elaborated in the paper. According to comparison research between the south and north part of east Tianshan area, the south part of east Tianshan is regarded as relative stable or relative weak in neotectonic movement in Neogene period and was selected as preselection area for high level radioactive waste repository. (authors)

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

    International Nuclear Information System (INIS)

    Kempe, T.F.; Martin, A.

    1981-05-01

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

  16. Development of a test system for high level liquid waste partitioning

    OpenAIRE

    Duan Wu H.; Chen Jing; Wang Jian C.; Wang Shu W.; Wang Xing H.

    2015-01-01

    The partitioning and transmutation strategy has increasingly attracted interest for the safe treatment and disposal of high level liquid waste, in which the partitioning of high level liquid waste is one of the critical technical issues. An improved total partitioning process, including a tri-alkylphosphine oxide process for the removal of actinides, a crown ether strontium extraction process for the removal of strontium, and a calixcrown ether cesium extra...

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

    International Nuclear Information System (INIS)

    Calmus, R.B.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  19. Development of an improved ion-exchange process for removing cesium and strontium from high-level radioactive liquid wastes

    International Nuclear Information System (INIS)

    Wallace, R.M.; Ferguson, R.B.

    1980-11-01

    Processes are being developed to solidify and isolate the biologically hazardous radionuclides from approximately 23 million gallons of alkaline high-level waste accumulated at the Savannah River Plant. The waste consists mainly of a liquid supernate, a damp salt cake, and a gelatinous, insoluble sludge. The reference solidification process involves separation of the water soluble fraction (supernate) from the insoluble fraction, removal of cesium and traces of strontium from the supernate, incorporation of the sludge and the radionuclides from the supernate in glass, and incorporation of the residual salt in concrete. A new process, now being developed, involves sorbing cesium on phenolic resins that contain no strongly acidic sulfonate groups. These resins can then be eluted with formic acid which is not possible with Duolite ARC-359. Duolite CS-100, a phenol-carboxylate resin, was chosen for further development because of its greater breakthrough capacity and because it also sorbs strontium to some extent. Strontium sorption by CS-100 was not sufficient to eliminate the need for Amberlite IRC-718. However, the latter resin can also be eluted with formic acid because its functional groups are weakly acidic. Formic acid elution permits several options to be considered. The preferred option consists simply of mixing the eluate with sludge prior to calcination. Sodium formate, which is formed when the resins in the sodium form are eluted, decomposes rapidly between 450 0 C and 500 0 C and will be destroyed in either the calciner or the melter. The resulting sodium oxide would be incorporated into glass. The principal advantage of the new process is the elimination of a number of process steps

  20. Removal of acid blue 062 on aqueous solution using calcinated colemanite ore waste

    Energy Technology Data Exchange (ETDEWEB)

    Atar, Necip [Department of Chemistry, Faculty of Arts and Science, University of Dumlupinar, Kuetahya (Turkey); Olgun, Asim [Department of Chemistry, Faculty of Arts and Science, University of Dumlupinar, Kuetahya (Turkey)]. E-mail: aolgun@dumlupinar.edu.tr

    2007-07-19

    Colemanite ore waste (CW) has been employed as adsorbent for the removal of acid blue 062 anionic dye (AB 062) from aqueous solution. The adsorption of AB 062 onto CW was examined with respect to contact time, calcination temperature, particle size, pH, adsorbent dosage and temperature. The physical and chemical properties of the CW, such as particle sizes and calcinations temperature, play important roles in dye adsorption. The dye adsorption largely depends on the initial pH of the solution with maximum uptake occurring at pH 1.Three simplified kinetics models, namely, pseudo-first order, pseudo-second order, and intraparticle diffusion models were tested to investigate the adsorption mechanisms. The kinetic adsorption of AB 062 on CW follows a pseudo-second order equation. The adsorption data have been analyzed using Langmuir and Freundlich isotherms. The results indicate that the Langmuir model provides the best correlation of the experimental data. Isotherms have also been used to obtain the thermodynamic parameters such as free energy, enthalpy and entropy of the adsorption of dye onto CW.

  1. Removal of acid blue 062 on aqueous solution using calcinated colemanite ore waste

    International Nuclear Information System (INIS)

    Atar, Necip; Olgun, Asim

    2007-01-01

    Colemanite ore waste (CW) has been employed as adsorbent for the removal of acid blue 062 anionic dye (AB 062) from aqueous solution. The adsorption of AB 062 onto CW was examined with respect to contact time, calcination temperature, particle size, pH, adsorbent dosage and temperature. The physical and chemical properties of the CW, such as particle sizes and calcinations temperature, play important roles in dye adsorption. The dye adsorption largely depends on the initial pH of the solution with maximum uptake occurring at pH 1.Three simplified kinetics models, namely, pseudo-first order, pseudo-second order, and intraparticle diffusion models were tested to investigate the adsorption mechanisms. The kinetic adsorption of AB 062 on CW follows a pseudo-second order equation. The adsorption data have been analyzed using Langmuir and Freundlich isotherms. The results indicate that the Langmuir model provides the best correlation of the experimental data. Isotherms have also been used to obtain the thermodynamic parameters such as free energy, enthalpy and entropy of the adsorption of dye onto CW

  2. Impact of partitioning and transmutation on high-level waste disposal for the fast breeder reactor fuel cycle

    International Nuclear Information System (INIS)

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

    2010-01-01

    The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large

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

    International Nuclear Information System (INIS)

    Cornman, W.R.

    1981-08-01

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

  4. Partitioning of high level liquid waste: experiences in plant level adoption

    International Nuclear Information System (INIS)

    Manohar, Smitha; Kaushik, C.P.

    2016-01-01

    High Level Radioactive Wastes are presently vitrified in borosilicate matrices in all our back end facilities in our country. This is in accordance with internationally endorsed methodology for the safe management of high level radioactive wastes. Recent advancements in the field of partitioning technology in our group, has presented us with an opportunity to have a fresh perspective on management of high level liquid radioactive wastes streams, that emanate from reprocessing operations. This paper will highlight our experiences with respect to both partitioning studies and vitrification practices, with a focus on waste volume reduction for final disposal. Incorporation of this technique has led to the implementation of the concept of recovering wealth from waste, a marked decrease on the load of disposal in deep geological repositories and serve as a step towards the vision of transmutation of long lived radionuclides

  5. Status of the high-level nuclear waste disposal program in Japan

    International Nuclear Information System (INIS)

    Uematsu, K.

    1985-01-01

    The Japan Atomic Energy Commission (JAEC) initiated a high-level radioactive waste disposal program in 1976. Since then, the Advisory Committee on Radioactive Waste Management of JAEC has revised the program twice. The latest revision was issued in 1984. The committee recommended a four-phase program and the last phase calls for the beginning of emplacement of the high-level nuclear waste into a selected repository in the Year 2000. The first phase is already completed, and the second phase of this decade calls for the selection of a candidate disposal site and the conducting of the RandD of waste disposal in an underground research laboratory and in a hot test facility. This paper covers the current status of the high-level nuclear waste disposal program in Japan

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

    International Nuclear Information System (INIS)

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

    1989-01-01

    This patent describes lead-iron phosphate glasses containing a high level of Fe 2 O 3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90 0 C, with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10 2 to 10 3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe 2 O 3 in forming the lead-iron phosphate glass is critical. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms

  7. Future directions of defense programs high-level waste technology programs

    International Nuclear Information System (INIS)

    Chee, T.C.; Shupe, M.W.; Turner, D.A.; Campbell, M.H.

    1987-01-01

    The Department of Energy has been managing high-level waste from the production of nuclear materials for defense activities over the last forty years. An objective for the Defense Waste and Transportation Management program is to develop technology which ensures the safe, permanent disposal of all defense radioactive wastes. Technology programs are underway to address the long-term strategy for permanent disposal of high-level waste generated at each Department of Energy site. Technology is being developed for assessing the hazards, environmental impacts, and costs of each long-term disposal alternative for selection and implementation. This paper addresses key technology development areas, and consideration of recent regulatory requirements associated with the long-term management of defense radioactive high-level waste

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    1991-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ray, J.W. [Savannah River Remediation (United States); Marra, S.L.; Herman, C.C. [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)

    2013-07-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    EIBLING, RUSSELLE.

    2004-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-10-01

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  15. The solidification of high-level liquid wastes in glass and ceramics

    International Nuclear Information System (INIS)

    Krause, H.

    1989-01-01

    In spent nuclear fuel reprocessing a highly radioactive waste solution is produced. It must be converted into a solid product, which binds the radionuclides, be hydrolytic as well as radiation and temperature resistant. Borosilicate glasses fulfil these requirements and, jointly with the barriers of a repository, they prevent inadmissible amounts of radionuclides from escaping into the biocycle. Two techniques were developed for industrial-scale vitrification: a rotary kiln calciner combined with an induction heated metallic melter and the electrode heated ceramic melters. Both techniques were already demonstrated on an industrial scale and under radioactive conditions. (AVM, Marcoule and PAMELA, Mol). (orig./MM) [de

  16. Small-scale demonstration of high-level radioactive waste processing and solidification using actual SRP waste

    International Nuclear Information System (INIS)

    Okeson, J.K.; Galloway, R.M.; Wilhite, E.L.; Woolsey, G.B.; Ferguson, R.B.

    1980-01-01

    A small-scale demonstration of the high-level radioactive waste solidification process by vitrification in borosilicate glass is being conducted using 5-6 liter batches of actual waste. Equipment performance and processing characteristics of the various unit operations in the process are reported and, where appropriate, are compared to large-scale results obtained with synthetic waste

  17. Small-scale integrated demonstration of high-level radioactive waste processing and vitrification using actual SRP waste

    International Nuclear Information System (INIS)

    Ferguson, R.B.; Woolsey, G.B.; Galloway, R.M.; Baumgarten, P.M.; Eibling, R.E.

    1980-01-01

    Experiments have been made to demonstrate the feasibility of immobilizing SRP high-level waste in borosilicate glass. Results to date are encouraging. Equipment performance and processing characteristics for solidifying small batches of actual SRP waste have agreed well with previous experience with small- and large-scale tests synthetic waste, and with theoretical predictions

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

    International Nuclear Information System (INIS)

    1995-01-01

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

  19. Economic considerations/comparisons for the disposal of defense high-level waste

    International Nuclear Information System (INIS)

    Leclaire, D.B.; Lazur, E.G.

    1985-01-01

    This paper provides a summary, in a generic sense, of the economic considerations and comparisons of permanent isolation of defense high-level waste (DHLW) in a licensed geologic repository. Topics considered include underground disposal, economic analysis, comparative evaluations, national defense, radioactive waste facilities, and licensing

  20. Safety principles and technical criteria for the underground disposal of high level radioactive wastes

    International Nuclear Information System (INIS)

    1989-01-01

    The main objective of this book is to set out an internationally agreed set of principles and criteria for the design of deep underground repositories for the disposal of high level radioactive wastes. This book is concerned with the post-closure period. Consideration of the operational requirements which must be met when wastes are being handled, stored and emplaced are not therefore included

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

    International Nuclear Information System (INIS)

    Anderson, P.A.

    1978-01-01

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

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

    Science.gov (United States)

    Hoffman, Darleane C.; Choppin, Gregory R.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    1984-10-01

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

  4. Stabilization of high-level waste from a chloride volatility nuclear fuel reprocessing system

    International Nuclear Information System (INIS)

    Smith, L.A.; Thornton, T.A.

    1979-01-01

    Methods for stabilizing high-level waste from a chloride volatility thorium-based fuel coprocessing system have been studied. The waste, which is present as chloride salts, is combined with SiO 2 or Al 2 O 3 and pyrohydrolyzed to remove the chloride ions. The resulting solid is then combined with a flux and glassified. 3 figures, 4 tables

  5. The development of a strategy for the management of high level radioactive wastes

    International Nuclear Information System (INIS)

    Beale, H.

    1981-07-01

    An assessment is made of the options available for the management of high level radioactive wastes. This preliminary study leads to the conclusion that the high active liquor should be vitrified at the earliest possible date and points to the advantages of storing the vitrified waste for an extended period in reinforced concrete casks. (author)

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

    International Nuclear Information System (INIS)

    Cornman, W.R.

    1981-02-01

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

  7. The scope and nature of the problem of high level nuclear waste disposal

    International Nuclear Information System (INIS)

    Jennekens, J.

    1981-09-01

    The disposal of high level nuclear waste poses a challenge to the Canadian technical and scientific communities, but a much greater challenge to government and industry leaders who must convince the public that the so-called 'problem' can be resolved by a pragmatic approach utilizing existing skills and knowledge. This paper outlines the objectives of radioactive waste management, the quantities of high level waste expected to be produced by the Canadian nuclear power program, the regulatory process which will apply and the government initiatives which have been and will be taken to ensure that the health, safety, security, and environmental interests of the public will be protected. (author)

  8. Current status of high level radioactive waste disposal in Japan and foreign countries

    International Nuclear Information System (INIS)

    Tanaka, Satoru; Tanabe, Hiromi; Inagaki, Yusuke; Ishida, Hisahiro; Kato, Osamu; Kurata, Mitsuyuki; Yamachika, Hidehiko

    2002-01-01

    At a time point of 2002, there is no country actually disposing high level radioactive wastes into grounds, but in most of countries legislative preparation and practicing agents are carried out and site selection is promoted together with energetic advancement of its R and Ds. As disposal methods of the high level radioactive wastes, various methods such as space disposal, oceanic bottom disposal, ice bed disposal, ground disposal, and so on have been examined. And, a processing technology called partitioning and transmutation technology separating long-lived radionuclides from liquid high level radioactive waste and transmutation into short-lived or harmless radionuclides has also been studied. Here was introduced their wrestling conditions in Japan and main foreign countries, as a special issue of the Current status of high level radioactive waste disposal in Japan and foreign countries'. The high level radioactive wastes (glassification solids or spent nuclear fuels) are wastes always formed by nuclear power generation and establishment of technologies is an important subject for nuclear fuel cycle. (G.K.)

  9. Proposed classification scheme for high-level and other radioactive wastes

    International Nuclear Information System (INIS)

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

    1986-01-01

    The Nuclear Waste Policy Act (NWPA) of 1982 defines high-level radioactive waste (HLW) as: (A) the highly radioactive material resulting from the reprocessing of spent nuclear fuel....that contains fission products in sufficient concentrations; and (B) other highly radioactive material that the Commission....determines....requires permanent isolation. This paper presents a generally applicable quantitative definition of HLW that addresses the description in paragraph (B). The approach also results in definitions of other waste classes, i.e., transuranic (TRU) and low-level waste (LLW). A basic waste classification scheme results from the quantitative definitions

  10. Evaluation of forms for the immobilization of high-level and transuranic wastes

    International Nuclear Information System (INIS)

    Schuman, R.P.; Cox, N.D.; Gibson, G.W.; Kelsey, P.V. Jr.

    1982-08-01

    A figure-of-merit (FOM) analysis has been made of a number of waste forms for solidifying both defense and commercial high-level reprocessing waste (HLW) and transuranic (TRU) wastes. The evaluation includes iron-enriched basalt (IEB), a fusion-produced glass-ceramic, which has not been included in other assessments. For HLW, concrete receives the highest FOM, but may not meet regulatory requirements; IEB and glass are the best choices of the materials that should easily meet regulatory requirements. Concrete waste forms are the best choice for TRU wastes, with IEB a close contender. 116 references, 3 figures, 112 tables

  11. A truck cask design for shipping defense high-level waste

    International Nuclear Information System (INIS)

    Madsen, M.M.; Zimmer, A.

    1985-01-01

    The Defense High-Level Waste (DHLW) cask is a Type B packaging currently under development by the U.S. Department of Energy (DOE). This truck cask has been designed to initially transport borosilicate glass waste from the Defense Waste Processing Facility (DWPF) to the Waste Isolation Pilot Plant (WIPP). Specific program activities include designing, testing, certifying, and fabricating a prototype legal-weight truck cask system. The design includes such state-of-the-art features as integral impact limiters and remote handling features. A replaceable shielding liner provides the flexibility for shipping a wide range of waste types and activity levels

  12. Alternatives Generation and Analysis for Phase 1 High-Level Waste Feed Tanks Selection

    International Nuclear Information System (INIS)

    CRAWFORD, T.W.

    1999-01-01

    A recent revision of the US Department of Energy privatization contract for the immobilization of high-level waste (HLW) at Hanford necessitates the investigation of alternative waste feed sources to meet contractual feed requirements. This analysis identifies wastes to be considered as HLW feeds and develops and conducts alternative analyses to comply with established criteria. A total of 12,426 cases involving 72 waste streams are evaluated and ranked in three cost-based alternative models. Additional programmatic criteria are assessed against leading alternative options to yield an optimum blended waste feed stream

  13. The geochemistry of high-level waste disposal in granitic rocks

    International Nuclear Information System (INIS)

    Chapman, N.A.; Sargent, F.P.

    1984-01-01

    Under the auspices of the cooperative agreement between Euratom and Atomic Energy of Canada Ltd about radioactive waste management and disposal, a joint workshop was held on the topic of the geochemistry of high-level waste disposal in granitic rocks. The report covers (1) waste form leaching, (2) thermodynamics, (3) geochemical models, (4) the role of colloids, (5) sorption phenomena, (6) the linking of flow and geochemical models, (7) microbial activity

  14. Integrating the commercial and defense high level waste programs - A utility perspective

    International Nuclear Information System (INIS)

    Tomonto, J.R.

    1986-01-01

    The Nuclear Waste Policy Act of 1982 provided that disposal of high-level wastes resulting from defense activities be included in the authorized repository unless the President determined that separate facilities are required. President Reagan approved commingling of defense and civilian wastes on April 30, 1985. The impacts of this decision on the repository schedule, civilian spent fuel acceptance rates, and cost sharing are reviewed and recommendations for resolving these issues are presented

  15. High-level waste solidification system for the Western New York Nuclear Service Center

    International Nuclear Information System (INIS)

    Carrell, J.R.; Holton, L.K.; Siemens, D.H.

    1982-01-01

    A preconceptual design for a waste conditioning and solidification system for the immobilization of the high-level liquid wastes (HLLW) stored at the Western New York Nuclear Service Center (WNYNSC), West Valley, New York was completed in 1981. The preconceptual design was conducted as part of the Department of Energy's (DOE) West Valley Demonstration Project, which requires a waste management demonstration at the WNYNSC. This paper summarizes the bases, assumptions, results and conclusions of the preconceptual design study

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

    Energy Technology Data Exchange (ETDEWEB)

    ROGERS, C.A.

    2000-02-17

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

  17. High-level-waste containment for a thousand years: unique technical and research problems

    International Nuclear Information System (INIS)

    Davis, M.S.

    1982-01-01

    In the United States the present policy for disposal of high level nuclear wastes is focused on isolation of solidified wastes in a mined geologic repository. Safe isolation is to be achieved by utilizing both natural and man-made barriers which will act in concert to assure the overall conservative performance of the disposal system. The incorporation of predictable man-made barriers into the waste disposal strategy has generated some new and unique problems for the scientific community

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

    International Nuclear Information System (INIS)

    Pavel Hrma; Pert Schill; Lubomir Nemec

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    ROGERS, C.A.

    2000-01-01

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

  20. Argentine project for the final disposal of high-level radioactive wastes

    International Nuclear Information System (INIS)

    Palacios, E.; Ciallella, N.R.; Petraitis, E.J.

    1989-01-01

    From 1980 Argentina is carrying out a research program on the final disposal of high level radioactive wastes. The quantity of wastes produced will be significant in next century. However, it was decided to start with the studies well in advance in order to demonstrate that the high level wastes could be disposed in a safety way. The option of the direct disposal of irradiated fuel elements was discarded, not only by the energetic value of the plutonium, but also for ecological reasons. In fact, the presence of a total inventory of actinides in the non-processed fuel would imply a more important radiological impact than that caused if the plutonium is recycled to produce energy. The decision to solve the technological aspects connected with the elimination of high-level radioactive wastes well in advance, was made to avoid transfering the problem to future generations. This decision is based not only on technical evaluations but also on ethic premises. (Author)